ORAU Team

Dose Reconstruction Project for NIOSH

Technical Basis Document for the Savannah River Site

To Be Used for EEOICPA Dose Reconstructions

 

Document Number:

ORAUT-TKBS-0003

Effective Date: 07/15/2003

Revision No.: 00

Controlled Copy No.: _______

Page 1 of 188

Subject Expert: Edward D. Scalsky

Approval: Signature on File___________Date: 07/15/2003 Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File________Date: 07/15/2003 Richard E. Toohey, Project Director

Approval: Signature on File___________Date: 07/15/2003 James W. Neton, OCAS Health Science Administrator

Supersedes:

 None

TABLE OF CONTENTS

TABLE OF CONTENTS ............................................................................................................ 1

LIST OF TABLES .... ................................................................................................................ 3

LIST OF FIGURES ...................................................................................................7

RECORD OF ISSUE / REVISIONS............................................................................................ 8

GLOSSARY ................................................................................... ............................. 9

ACRONYMS AND ABBREVIATIONS......................................................................................16

EXECUTIVE SUMMARY ..........................................................................................................19

1.0 INTRODUCTION...........................................................................................................22

1.1 Purpose .............................................................................................................22

1.2 Scope ................................................................................................................22

1.3 Site Activities and Processes .............................................................................23

2.0 OCCUPATIONAL MEDICAL DOSE .............................................................................45

2.1 Examination Frequencies...................................................................................45

2.2 Equipment and Techniques ...............................................................................45

2.3 Organ Dose Calculations ...................................................................................47

2.4 Uncertainty.........................................................................................................49

3.0 OCCUPATIONAL ENVIRONMENTAL DOSE...............................................................51

3.1 Internal Dose from On-Site Atmospheric Radionuclide Concentrations..............51

3.1.1 On-Site Releases to Air......................................................................................51

3.1.2 Annual intakes from Resuspension....................................................................59

3.2 Occupational External Dose...............................................................................60

3.2.1 Ambient Radiation..............................................................................................60

3.2.2 Releases of Noble Gases ..................................................................................61
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 2 of 188

3.3 Uncertainty.........................................................................................................62

4.0 OCCUPATIONAL INTERNAL DOSE............................................................................63

4.1 In vitro Minimum Detectable Activities (MDAs), Counting Methods ....................63

4.1.1 In vitro Urine Analysis ........................................................................................64

4.1.2 Correcting for Urinalysis Volume........................................................................71

4.1.3 Fecal Sample Analysis.......................................................................................72

4.2 In vivo MDAs, Counting Methods, and Reporting Practices ...............................72

4.2.1 Whole Body Counting ........................................................................................72

4.2.2 Chest Counting ..................................................................................................73

4.3 Personal Air Sampling Data as Applicable .........................................................75

4.4 Interferences, Uncertainty, and Use of Reporting Levels....................................75

4.4.1 Contamination of Samples .................................................................................75

4.4.2 Uncertainty.........................................................................................................76

4.4.3 Use of Reporting Levels and Less-Than Values ................................................76

5.0 OCCUPATIONAL EXTERNAL DOSIMETRY................................................................78

5.1 Introduction........................................................................................................78

5.2 Basis of Comparison..........................................................................................78

5.3 Dose Reconstruction Parameters ......................................................................79

5.3.1 SRS Historical Administrative Practices .............................................................80

5.3.2 SRS Dosimetry Technology ...............................................................................82

5.3.3 Calibration..........................................................................................................87

5.3.4 Workplace Radiation Fields ...............................................................................89

5.3.5 SRS Workplace Neutron Dose Uncertainty......................................................101

5.4 Adjustments to Recorded Dose........................................................................103

5.4.1 Photon Dose Adjustments................................................................................103

5.4.2 Neutron Dose Adjustments ..............................................................................103

5.5 Missed Dose....................................................................................................106

5.5.1 Photon Missed Dose........................................................................................106

5.5.2 Neutron Missed Dose ......................................................................................108

5.6 Organ Dose .....................................................................................................110

5.6.1 Organ Dose Conversion Factors......................................................................110

6.0 REFERENCES............................................................................................................111

ATTACHMENT A DESCRIPTION OF FACILITIES AND PROCESSES.............................123

ATTACHMENT B OCCUPATIONAL MEDICAL DOSE ......................................................143

ATTACHMENT D OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS ..172

ATTACHMENT E OCCUPATIONAL EXTERNAL DOSE FOR MONITORED WORKERS . 178
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 3 of 188

LIST OF TABLES

Table 2.01 Frequency of Occupational Chest X-rays at the SRS .............................46

Table 2.02 Description of the X-ray Equipment used at SRS ...................................47

Table 2.03 Technique factors Used for Each Type of X-ray Equipment....................47

Table 2.04 Analogues for IREP Organs Not Included in ICRP 34.............................49

Table 3.1.1-1 Distance, in Meters, from Source Areas to Population Centers at SRS. .53

Table 3.1.1-2 Overview of Tritium Monitors at SRS (Table E-1, CDC 2001, p. E-2) .....56

Table 3.1.1-3 Overview of Radioiodine Sampling and Monitoring Methods (RAC Report 

(CDC 2001, Table 4.2-2)).......................................................................57

Table 3.1.2-1 Radionuclide Soil Concentrations, pCi g-1, (1 sigma) Dry Weight (0-8 cm 

depth).....................................................................................................60

Table 4.1.1-1 Current MDAs for Urinalysis ...................................................................64

Table 4.1.1-2 Pu-239/240 – in vitro Urinalysis ..............................................................66

Table 4.1.1-3 Pu-238 – in vitro Urinalysis.....................................................................66

Table 4.1.1-4 Activity Composition for Reference 6% and 12% 240Pu Mixturesa ...........67

Table 4.1.1-5 Uranium Urinalysis .................................................................................69

Table 4.2.1-1 Whole Body Counting.............................................................................73

Table 4.2.1-2 Current MDAs for Whole Body Counting ................................................73

Table 4.2.2-1 Current MDAs for Chest Counting (2.5 cm chest wall thickness)............74

Table 4.2.2-2 Current MDAs for Chest Counting 50th Percentile and 95th Percentile Male 

Workers..................................................................................................75

Table 4.4.3-1 239Pu Chronic Inhalation Intake Assessment Based on MDA Urinalysis on 

Last Day of the Period............................................................................77

Table 5.3.1-1 SRS Dosimeter Type, Period of Use, Exchange Frequency, MDL and 

Potential Annual Missed Dose................................................................81

Table 5.3.2-1 SRS Historical Dosimetry Events (Taylor et al 1995)..............................83

Table 5.3.2.1-1 IARC Testing Results for US beta/photon Dosimeters............................85

Table 5.3.2.1-2 Testing Results for Hanford Two-Element and Multi-Element Film 

Dosimeters for Energy and Angular Response(a,b) ..................................86

Table 5.3.3.1-1 Common Sources of Laboratory Bias for Beta/photon Dosimeter 

Calibration Parameters(a) ........................................................................88

Table 5.3.3.2-1 Common Sources of Laboratory Bias for Neutron Dosimeter Calibration 

Parameters(a)..........................................................................................89

Table 5.3.4.1-1 Selection of Beta and Photon Radiation Energies and Percentages for 

SRS Facilities.........................................................................................90

Table 5.3.4.1-2 Common Workplace Beta/Photon Dosimeter Hp(10) Performance(a).......91
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 4 of 188

Table 5.3.4.2.1.1-1 Measured Dose Fractions at Different Locations in H Area ....................93

Table 5.3.4.2.1.1-2 Measured Dose Fractions at Different Locations in F Area.....................94

Table 5.3.4.2.1.2-1 Recording Periods for Selected Hanford Plutonium Workers..................95

Table 5.3.4.2.1.2-2 Hanford Ratio of Recorded Dose Components.......................................96

Table 5.3.4.2.1.1-3 Post 1971 SRS Ratio of FB and HB Line Recorded Dose Components .96

Table 5.3.4.2.2.1-1 Dose Fractions for Various Operations in 773-A Area ............................97

Table 5.3.4.2.2.2-1 Post 1971 SRS Ratio of 773-A Recorded Dose Components.................97

Table 5.3.4.2.3.1-1 Dose Fractions for Heavy Water Reactors at SRS .................................99

Table 5.3.4.2.3.2-1 Post 1971 SRS Ratio of Plutonium and Reactor Recorded Dose 

Components...........................................................................................99

Table 5.3.4.2.3.1-1 Dose Fractions at the Calibration Facility (736-A)................................. 100

Table 5.3.4.2-3 Common Workplace Neutron Dosimeter Hp(10) Performance(a).........102

Table 5.4.1-1 Adjustments to Reported SRS Deep Photon Dose...............................103

Table 5.4.2-1 Historical Neutron Quality or Weighting Factors ...................................104

Table 5.4.2.2-1 Summary of Dose Fractions and associated ICRP 60 correction factors

.............................................................................................................105

Table 5.5.1-1 Missed Photon Dose According to SRS Facility....................................106

Table 5.5.1-2 Missed Photon Dose According to Dosimeter Type..............................107

Table 5.5.1-3 Missed Photon Dose Adjustments to Recorded Deep Dose .................107

Table 5.5.2-1 Summary of SRS Neutron Dosimeter Limits of Detection (LOD) ..........108

Table 5.5.2-2 Summary of neutron to photon ratios by process..................................109

Table 5.6.1-1 Default Exposure Geometries to calculate Organ Dose........................110

Table A-1.1 Reactors Years of Operation ................................................................123

Table A-1.2 Radionuclides of Concern for All Reactors (a)........................................123

Table A-2 F Area A-Line Facility ...........................................................................124

Table A-3 221-F B-Line Facility ............................................................................124

Table A-4.1 221-F Canyon Facility...........................................................................125

Table A-4.2 221-F Canyon Facility...........................................................................125

Table A-4.3 221-F Canyon Facility...........................................................................126

Table A-5 New Special Recovery (NSR) Facility...................................................126

Table A-6.1 F-Area Outside Facilities ......................................................................127

Table A-6.2 F Area Outside Facilities ......................................................................127

Table A-6.3 F Area Outside Facilities ......................................................................128

Table A-6.4 F Area Outside Facilities ......................................................................128

Table A-6.5 F Area Outside Facilities Table.............................................................129
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 5 of 188

Table A-6.6 F Area Outside Facilities ......................................................................129

Table A-7 238PuO2 Fuel Form Facility (PuFF) and 238PuO2 Experimental Facility 

(PEF)....................................................................................................130

Table A-8 235-F Vaults.........................................................................................130

Table A-9 772-F and 772-1F Production Control Laboratories.............................131

Table A-10 E-Area Solid Waste Disposal Facility (SWDF)......................................131

Table A-11 Plutonium Storage Facility (PSF)..........................................................131

Table A-12 F/H Effluent Treatment Facility (ETF) ...................................................132

Table A-13 F & H Area - Cooling Water and Retention Basins ...............................132

Table A-14 F & H Area Tank Farms........................................................................133

Table A-15 Waste Certification Facility (WCF)........................................................133

Table A-16.1 221-H B-Line Facility ............................................................................134

Table A-16.2 221-H B-Line Facility ............................................................................134

Table A-16.3 221-H B-Line Facility. Np02 andPu02 Facility ........................................134

Table A-17.1 H-Canyon Facility .................................................................................135

Table A-17.2 H-Canyon Facility .................................................................................135

Table A-17.3 H-Canyon Facility .................................................................................136

Table A-17.4 H-Canyon Facility .................................................................................136

Table A-18.1 H-Area Outside Facilities......................................................................137

Table A-18.2 H - Area Outside Facilities....................................................................137

Table A-19.1 Receiving Basin for Off-site Fuel (RBOF) .............................................138

Table A-19.2 Resin Regeneration Facility (RRF) .......................................................138

Table A-20 Tritium Facilities (a)................................................................................138

Table A-21 Uranium Target Fabrication Facility, 313 M ..........................................139

Table A-22 Uranium Processing Facilities 320-M, 322-M, and 341-M (a) .................139

Table A-23 Fuel Fabrication Facility, 321-M (a) ........................................................140

Table A-24.1 S Area Defense Waste Processing Facility (DWPF).............................140

Table A-24.2 S Area Defense waste Processing Facility (DWPF)..............................141

Table A-25.1 Z Area ..................................................................................................141

Table A-25.2 Z Area ..................................................................................................142

Table A-26 A Area ..................................................................................................142

Table B-01 Doses for Organs Identified in ICRP 34 (1982) Beam Quality for 3.5 mm 

Al HVL..................................................................................................144

Table B-02 Dose For IREP Organs Not Included in ICRP 34 (1982) Beam Quality for 

3.5 mm Al HVL .....................................................................................145
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 6 of 188

Table C-01 A Area Annual Intakes and geometric standard deviations, (Bq per year)

.............................................................................................................147

Table C-02 C Reactor Area Annual Intakes and geometric standard deviations, (Bq 

per year)...............................................................................................148

Table C-03 D Area Annual Intakes and geometric standard deviations, (Bq per year)

.............................................................................................................149

Table C-04 E Area Annual Intakes and geometric standard deviations, (Bq per year)

.............................................................................................................150

Table C-05 F Area Annual Intakes and geometric standard deviations, (Bq per year)

.............................................................................................................151

Table C-06 H Area Annual Intakes and geometric standard deviations, (Bq per year)

.............................................................................................................152

Table C-07 S Area Annual Intakes and geometric standard deviations, (Bq per year)

.............................................................................................................153

Table C-08 Z Area Annual Intakes and geometric standard deviations, (Bq per year)

.............................................................................................................154

Table C-09 Forest Service Headquarters (a) Calculated Air Concentration (Bq per m3

)

.............................................................................................................155

Table C-10 Central Shops Annual Intakes and geometric standard deviations, (Bq per 

year).....................................................................................................156

Table C-11 L Reactor Area Annual Intakes and geometric standard deviations, (Bq 

per year)...............................................................................................157

Table C-12 P Reactor Area Annual Intakes and geometric standard deviations, (Bq 

per year)...............................................................................................158

Table C-13 K Reactor Area Annual Intakes and geometric standard deviations, (Bq 

per year)...............................................................................................159

Table C-14 R Reactor Area Annual Intakes and geometric standard deviations, (Bq 

per year)...............................................................................................160

Table C-15 Rail Yard Annual Intakes and geometric standard deviations, (Bq per 

year).....................................................................................................161

Table C-16 Average Site-Wide (a) Annual Intakes and Geometric Standard Deviations, 

(Bq per year) ........................................................................................162

Table C-17 Maximum Site-Wide Annual Intakes and geometric standard deviations, 

(Bq per year) ........................................................................................163

Table C-18 Annual Intakes for F and H Areas (Bq per year) ...................................164

Table C-19 Ambient radiation levels for selected SRS locations (mrem/year).........165

Table C-20 Argon-41 50th Percentile Dose(a) to Skin(b) and Bone Surfaces(c) Resulting 

From Reactor Releases (mrem per year) and geometric standard 

deviation...............................................................................................168
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 7 of 188

Table C-21 Argon-41 50th Percentile Dose(a) (b) to Whole Body and Other Organs 

Resulting From Reactor Releases (mrem per year) and geometric 

standard deviation ................................................................................170

Table D-01 Limits of Detection for Urinalysisa .........................................................174

Table D-02 MDAs for Whole Body Counting ...........................................................176

Table D-03 MDAs for Chest Counting.....................................................................177

Table E-01 IREP Dose Parameter Input Screen.....................................................178

Table E-02 Selection of Beta and Photon Radiation Energies and Percentages for 

SRS Facilities and Processes...............................................................179

Table E-03 Selection of Neutron Radiation Energies and Percentages for SRS 

Facilities and Processes.......................................................................180

Table E-04 Adjustments to Reported SRS Deep Photon Dose...............................181

Table E-05 Summary of neutron Dose Fractions and associated ICRP 60 correction 

factors ..................................................................................................183

Table E-06 Missed photon dose according to SRS Facility.....................................184

Table E-07 Missed Photon Dose According to Dosimeter Type..............................184

Table E-08 Missed Photon Dose According to Year ...............................................185

Table E-09 Neutron to Photon Ratios by SRS Process...........................................186

Table E-10 SRS TLND Limits of Detection .............................................................187

Table E-11 Default Exposure Geometries to Calculate Organ Dose.......................187

LIST OF FIGURES

Figure 3.1.1-1 The SRS showing the reactor (C-, K-, L-, P-. and R-Reactors), and 

processing areas (F Area and H Area) in a rough circle toward the center 

of the Site, and D Area and A and M Areas near the Site perimeter. Also 

shown are DWPF (S Area), Saltstone (Z Area), Central Shops, Heavy 

Water Extraction (D Area), U.S. Forest Service Headquarters, and the 

Railroad Classification Yard....................................................................52

Figure 5.3.1-1 SRS Monitored Workers and Collective Dose ........................................82

Figure 5.3.2.1-1 Estimated SRS Dosimeter Photon Response Characteristics.................85

Figure 5.3.4.2-1 Neutron Energy Spectra Recorded at SRS Plutonium Facilities with 

Neutron Energy Groups Overlay. ...........................................................92

Figure 5.3.4.2-2 Trends in SRS and Hanford Collective Neutron Dose Normalized to 

Plutonium Production. ............................................................................95

Figure 5.3.4.2-3 Neutron Energy Spectra Recorded Behind the K-Reactor Shield Door 

with Neutron Energy Groups Overlay. ....................................................98

Figure 5.3.4.2-4 Comparison of Various Neutron Spectra from different Plutonium 

Compounds and Alloys (Brackenbush et al, 1987) ...............................101
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 8 of 188

RECORD OF ISSUE / REVISIONS

ISSUE

AUTHORIIZATION

DATE

EFFECTIVE 

DATE

REV.

NO.

 DESCRIPTION

Draft 05/10/2003 00-A Original issue draft technical basis document for the 

Savannah River Site. Initiated by Judson L. 

Kenoyer.

Draft 05/19/2003 00-B Revised to correct formatting errors and make 

additional text changes. Initiated by Judson L. 

Kenoyer.

Draft 07/03/2003 00-C Revised to incorporate changes based on 

information gathered through contacts with SRS 

technical subject matter experts. Initiated by 

Edward D. Scalsky.

Draft 07/14/2003 00-D Revised to respond to comments provided by 

NIOSH on July 11, 2003. Initiated by Edward D. 

Scalsky.

First Approved 

Issue

07/15/2003 00 Initiated by Edward D. Scalsky.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 9 of 188

GLOSSARY

“A,” “B,” “C,” and “D”: codes used in whole body count results; the letters represent 

energy regions: “A” means 810-910keV, “B” means 920-1020 keV, “C” means 1180-1290 

keV, and “D” means 1760-1990 keV.

 

absorption type: a term introduced in ICRP’s revised respiratory tract model (ICRP66) to replace the clearance classes in ICRP-30. The absorption types are Type F 

for fast absorption, formerly called Class D; Type M for moderate absorption, 

formerly called Class W; and Type S for slow absorption, formerly called Class Y. 

Also referred to as Type. 

albedo dosimeter: A dosimeter that measures the slow neutrons which are generated 

by higher energy neutrons incident on the body and which reflect back into the 

dosimeter. 

annual dose equivalent: the dose equivalent received in a year. The annual dose 

equivalent is expressed in units of rem (sievert).

Atomic Energy Commission: original agency established for nuclear weapons and power 

production; a predecessor to the U.S. Department of Energy.

beta (β) dose: a designation (i.e., beta) on some SRS external dose records referring to 

the dose from less-energetic beta, X ray and/or gamma radiation (see ALSO open window, or 

shallow dose). 

beta radiation: radiation consisting of charged particles of very small mass (i.e., the 

electron) emitted spontaneously from the nuclei of certain radioactive elements. Most (if not 

all) of the direct fission products emit beta radiation. Physically, the beta particle is identical 

to an electron moving at high velocity.

class: The respiratory tract inhalation classification scheme developed in ICRP-30 for 

inhaled material according to its rate of clearance from the pulmonary region of the 

respiratory tract. Materials are classified as D (days), W (weeks), or Y (years), according to 

how fast they clear the lungs; Class D in less than 10 days; Class W in 10 to 100 days; 

Class Y in more than 100 days. Recent recommendations in ICRP Report 66 replaced 

classes D, W, and Y with lung absorption Types F (fast), M (moderate), and S (slow).

collective dose equivalent: the sum of the dose equivalents of all individuals in an 

exposed population. Collective dose is expressed in units of person-rem (person-sievert).

collective effective dose equivalent: the sum of the effective dose equivalents of all 

individuals in an exposed population. Collective effective dose is expressed in units of 

person-rem (person-sievert).

curie: a special unit of activity. One curie exactly equals 3.7 x 1010 nuclear transitions per 

second.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 10 of 188

deep absorbed dose (Dd): the absorbed dose at the depth of 1.0 cm in a material of 

specified geometry and composition.

deep dose equivalent (Hd): the dose equivalent at the respective depth of 1.0 cm in tissue.

detection limit (lower): the minimum quantifiable exposure or neutron flux that can be 

detected.

delayed neutron analysis (DNA): the method was adopted for both enriched and depleted 

uranium urinalysis analyses in about 1982.

depleted uranium (DU) analysis: indicates that the chemical properties of uranium were 

used for analysis, not necessarily that the material in question was depleted uranium

dose equivalent (H): the product of the absorbed dose (D), the quality factor (Q), and any 

other modifying factors. The special unit is the rem. When D is expressed in Gy, H is in 

Sieverts (Sv). (1 Sv = 100 rem.)

dose of record: the dose files provided by DOE to NIOSH as part of the individual worker 

files. 

dosimeter: a device used to measure the quantity of radiation received. A holder with 

radiation-absorbing elements (filters) and an insert with radiation-sensitive elements 

packaged to provide a record of absorbed dose or dose equivalent received by an 

individual. (see albedo dosimeter, film dosimeter, neutron film dosimeter, 

thermoluminescent dosimeter.)

dosimetry: the science of assessing absorbed dose, dose equivalent, effective dose 

equivalent, etc., from external and/or internal sources of radiation. 

dosimetry system: a system used to assess dose equivalent from external radiation to the 

whole body, skin, and/or extremities. This includes the fabrication, assignment, and 

processing of the dosimeters as well as interpretation and documentation of the results.

DU: as used in this document depleted uranium, used as targets for 239Pu production; 

isotopic activity ratios (NOTE: this is not the mass ratio) listed by SRS as:

Isotope

234U 0.0840

Activity fraction

235U 0.0145

238U 0.9015

enriched uranium (EU) analysis: as used in this document indicates the analysis was by 

alpha counting, not necessarily that the material in question was enriched uranium

exchange period (frequency): time period (weekly, biweekly, monthly, quarterly, etc.) for 

routine exchange of dosimeters.

exposure: as used in the technical sense, exposure refers to a measure expressed in 

roentgens (R) of the ionization produced by photon radiation (i.e., gamma and X rays) in air. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 11 of 188

extremity: that portion of the arm extending from and including the elbow through the 

fingertips, and that portion of the leg extending from and including the knee and patella 

through the tips of the toes.

FB-Line: the plutonium finishing line located in the 221-F facility.

Field (f): refers to SRS practice to quantify Neutron Track Emulsion Type A (NTA) neutron 

tracks using a projection image covering a 15-inch projection screen. 

field calibration: dosimeter calibration based on radiation types, intensity and energies 

present in the work environment.

film: generally means a "film packet" that contains one or more pieces of film in a light-tight 

wrapping. The film when developed has an image caused by radiation that can be 

measured using an optical densitometer. (see Dupont 552, Dupont 558, Eastman Kodak, 

Nuclear Emulsions.)

film density: see optical density.

film dosimeter: a small packet of film within a holder that attaches to a wearer.

flux density (n/cm2

-sec): a measure of the intensity of neutron radiation in neutrons/cm2-

sec. It is the number of neutrons passing through 1 square centimeter of a given target in 1 

second. Expressed as nv, where n = the number of neutrons per cubic centimeter and v = 

their velocity in centimeters per second. 

gamma rays (G or γ): electromagnetic radiation (photons) originating in atomic nuclei and 

accompanying many nuclear reactions (e.g., fission, radioactive decay, and neutron 

capture). Physically, gamma rays are identical to X rays of high energy, the only essential 

difference being that X rays do not originate in the nucleus. 

Gray (Gy): The special name for the SI unit of absorbed dose. (1 Gy = 1 Jkg-1) 

HB-Line: the plutonium finishing line located in the 221-H facility.

HEU: highly enriched uranium, used in Naval Reactor Fuel; isotopic ratios listed by SRS as:

 Isotope 

234U 0.9806

Activity fraction

 235U 0.0194

238U 0.0000

Health Protection Radiation Exposure Database (HPRED): this database replaced the 

mainframe tape master file system in April 1989. This system interfaced with the TLD badge 

reader system. In 1990, visitor records were included in this database.

Health Protection Annual Radiation Exposure History Database (HPAREH): Starting in 

1980, this database was used to generate yearly radiation exposure reports to employees.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 12 of 188

ionizing radiation: electromagnetic or particulate radiation capable of producing charged 

particles through interactions with matter.

liquid scintillation counter (LSC): used for tritium urinalysis starting in 1958.

maximum permissible body burden (MPBB): limit for body content of a radionuclide used 

from startup until 1975

maximum permissible lung burden (MPLB): during the 1970's the occupational limit for 

Pu was expressed in terms of a quantity of plutonium that could be present in the chest at 

any given time, equal to 16 μCi (0.25 μg) 239Pu.

maximum permissible organ burden (MPOB): limit for organ content of a radionuclide 

used from startup until 1975.

neutron: a basic particle that is electrically neutral, having nearly the same mass as the 

hydrogen atom.

neutron, fast: neutrons with energy equal or greater than 10 keV.

neutron, intermediate: neutrons with energy between 0.5 eV and 10 keV.

neutron, thermal: strictly, neutrons in thermal equilibrium with surroundings. Generally, 

neutrons with energy less than about 0.5 eV.

neutron film dosimeter: a film dosimeter that contains an Neutron Track Emulsion, type A, 

film packet.

nuclear emulsion: often referred to as “NTA” film and used to measure personnel dose 

from neutron radiation.

Nuclear Track Emulsion, Type A (NTA): a film that is sensitive to fast neutrons. The 

developed image has tracks caused by neutrons that can be seen by using an appropriate 

imaging capability such as oil immersion and a 1000X power microscope or a projection 

capability.

open window (OW): designation on SRS film dosimeter reports that implies the use of little 

(i.e., only security credential) shielding. It commonly is used to label the film response 

corresponding to the open window area. 

operating area: designation of SRS major operational work areas among the respective 

fuel fabrication (300), reactor operations (R, P, L, K and C), chemical separations (F and H), 

uranium separation (A-Line), plutonium finishing (H-Canyon, B-Line), research and 

development, and transportation, communication and general site support.

optical density: the quantitative measurement of photographic blackening; the optical 

density is defined as D = Log10 (Io/I).
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 13 of 188

Parameter 1: the column in the IREP template where the dose reconstructor will enter the 

calculated dose. Multiple entries based on year of employment, type of radiation and 

appropriate energy ranges, internal and external exposure are possible.

Parameter 2: the column in the IREP template where the dose reconstructor will enter the 

lower limit of the dose distribution based on the radiation type and the dose distribution type.

PuF4 source: a neutron source whose activating material is plutonium fluoride. The source 

was used to duplicate the neutron energies in SRS plutonium handling facilities.

pencil dosimeters: a type of ionization chamber used by personnel to measure radiation 

dose. These results may be labeled as “Pen” dose. Other names: pencil, pocket 

dosimeter, pocket pencil, pocket ionization chamber (PIC).

personal dose equivalent Hp(d): represents the dose equivalent in soft tissue below a 

specified point on the body at an appropriate depth d. The depths selected for personnel 

dosimetry are 0.07 mm and 10 mm, respectively, for the skin and body. These are noted as 

Hp(0.07) and Hp(10), respectively. 

photon: a unit or "particle" of electromagnetic radiation consisting of x- and/or gamma rays. 

photon - X ray: electromagnetic radiation of energies between 10 keV and 100 keV whose 

source can be an X-ray machine or radionuclide.

quality factor, Q: a modifying factor used to derive dose equivalent from absorbed dose.

Rad: the unit of absorbed dose 

radiation: alpha, beta, neutron, and photon radiation. 

radioactivity: the spontaneous emission of radiation, generally alpha or beta particles, 

gamma rays, and neutrons from unstable nuclei.

Radiological Qualification Badge (RQB): System starting in 1992 to control each 

worker’s dosimetry requirements including bioassay requirements. Each radiation worker 

carries his/her personal card. From 1992 to 1999 the card listed requirements that had 

been met by the worker prospectively prior to entering radiation areas. That is, the card 

requirements and the Radiation Work Permit requirements had to match prior to entry. 

From 1999 to present the requirements on the card are listed retrospectively based on 

entries made during the designated period. Prior to the RQB system, an electronic 

database was used to track worker dosimetry requirements. 

relative biological effectiveness (RBE): a ratio of the absorbed dose of a reference 

radiation to the absorbed dose of a test radiation producing the same biological effects, 

other conditions being equal.

rem: the rem is a unit of dose equivalent, which is equal to the product of the number of rad 

absorbed and the "quality factor."
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 14 of 188

rep: a historical unit, roentgen-equivalent-physical (mrep = millirep) used when reporting 

beta exposures, usually recorded in mrep; equivalent to 93 ergs of energy per gram. 

Roentgen (R or r): a unit of exposure to gamma (or X ray) radiation. It is defined precisely 

as the quantity of gamma (or x) rays that will produce a total charge of 2.58 x 10-4 coulomb 

in 1 kg of dry air STP. An exposure of 1 R is approximately equivalent to an absorbed dose 

of 1 rad in soft tissue for higher (~>100 keV) energy photons.

RU: recycled uranium, is SRS reactor fuel; isotopic activity ratios listed by SRS as:

Isotope 234U 0.8489Activity fraction

 235U 0.0120

 236U 0.1388

 238U 0.0003 or 0.0004 (both listed)

shallow absorbed dose (Ds): the absorbed dose at a depth of 0.07 mm in a material of 

specified geometry and composition.

shallow dose equivalent (Hs): dose equivalent at a depth of 0.07 mm in tissue.

shielding: any material or obstruction that absorbs (or attenuates) radiation and thus tends 

to protect personnel or materials from radiation.

shield: designation on SRS dose records that imply a radiation dose, typically to the whole 

body, from higher energy photon radiation.

sievert (Sv): the SI unit for dose equivalent. (1 Sv = 100 rem.)

silver shield(s): the 1-mm thick shields covering the film packet in the early SRS personnel 

film dosimeters. 

skin dose: absorbed dose at a tissue depth of 7 mg/cm2 ~ 0.07 mm is tissue..

Snoopy: portable neutron monitoring instrument with a moderated BF3 detector.

SP: this historical designation stands for “Special Product” and referred to 237Np.

thermoluminescence: property of a material that causes it to emit light as a result of being 

excited by heat.

thermoluminescent dosimeter (TLD): a holder containing solid chips of material that 

when heated will release the stored energy as light. The measurement of this light provides 

a measurement of absorbed dose. 

3% Plutonium, 6% Plutonium, and 12% Plutonium: designations for typical mixtures of 

plutonium encountered at SRS, where the percentage represents the percent by mass of 

240Pu in the mixture.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 15 of 188

whole body dose: commonly defined as the absorbed dose at a tissue depth of 1.0 cm 

(1000 mg/cm2

); however, this term is also used to refer to the recorded dose.

X ray: ionizing electromagnetic radiation of external nuclear origin or a radiograph
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 16 of 188

ACRONYMS AND ABBREVIATIONS

α alpha particle

AEC (U.S.) Atomic Energy Commission

AED Atomic Energy Division (of SRP)

AECL Administrative Exposure Control Level

AEDE Annual Effective Dose Equivalent 

ARU Acid Recovery Unit

AWE Atomic Weapons Employer

β beta particle

BSRI Bechtel Savannah River Incorporated

BSRI Bechtel Savannah River Institute 

BTM Berthold Tritium Monitor

CATI Computer Assisted Telephone Interview 

CEDE Committed Effective Dose Equivalent (replaced the AEDE in 1992)

 c/m counts per minute

CND Criticality Neutron Dosimeter

DAC Derived Air Concentration

DAC-h Derived Air Concentration hours

DCF Dose Conversion Factor

DNA Delayed Neutron Analysis

DE Dose Equivalent

DL Detection Limit

DOE U. S. Department of Energy

DOELAP DOE Laboratory Accreditation Program 

DOL Department of Labor

DPSOP Du Pont Standard Operating Procedure 

DWPF Defense Waste Processing Facility

DU Depleted Uranium

EEOICPA Energy Employees Occupational Illness Compensation Program Act

ESE Entrance Skin Exposure

ERDA Energy Research & Development Administration

ETF Effluent Treatment Facility

γ gamma ray

HEU Highly Enriched Uranium

HHS Department of Health and Human Services

HHE High Heat Waste

HLW High Level Waste

Hp (0.07) Dose equivalent at 0.07 mm depth

Hp(10) Personal Dose Equivalent at 10 mm depth

HPRED Health Protection Radiation Exposure Database 

HPAREH Health Protection Annual Radiation Exposure History Database 

HRTM Human Respiratory Tract Model
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 17 of 188

HTO Tritiated Water

HVL Half Value Layer

IARC International Agency for Research on Cancer

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis 

IREP Interactive RadioEpidemiological Program

KPA Kinetic Phosphorimetry Analysis

LANL Los Alamos National Laboratory

LLD Lower Limit of Detection

LSC Liquid Scintillation Counter

LHW Low Heat Waste

MDA Minimum Detectable Activity

MDL Minimum Detection Limit

MED Manhattan Engineering District

MEPAS Multimedia Environmental Pollutant Assessment System

MPBB Maximum Permissible Body Burden

MPLB Maximum Permissible Lung Burden

MPOB Maximum Permissible Organ Burden

NCRP National Council on Radiation Protection and Measurements

NIOSH National Institute for Occupational Safety and Health

NOCTS NIOSH OCAS Claims Tracking System

NTA Nuclear Track Emulsion

OCAS (NIOSH) Office of Compensation Analysis and Support 

ORAU Oak Ridge Associated Universities

ORNL Oak Ridge National Laboratory

PA Posterior-Anterior (medical X ray)

PAS Personal Air Sampler

PDIS Personnel Dosimetry Intercomparison Studies

PEF 238PuO2 Experimental Facility

PIC Pocket Ionization Chamber (i.e., “Pencil” dosimeter)

PSF Plutonium Storage Facility

PuFF 238PuO2 Fuel Form Facility

PUREX Plutonium Uranium Extraction (Facility/Process)

R Roentgen

RAS Regional Air Sampler

RBE Relative Biological Effectiveness

RBOF Receiving Basin for Offsite Fuel

RC & HP Radiological Control and Health Physics Department 

rem Roentgen equivalent man

rep Roentgen equivalent physical

RMS Root Mean Square
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 18 of 188

RQB: Radiological Qualification Badge

RRF Resin Regeneration Facility

RTF Replacement Tritium Facility

RU Recycled Uranium

SEC Special Exposure Cohort

SSD Source to Skin Distance

SID Source to Image Distance

SMI Stack Monitor Integrator

SRI Savannah River Institute

SRL Savannah River Laboratory

SRO Savannah River Operations Office 

SRP Savannah River Plant

SRS Savannah River Site

STM Stack Tritium Monitor

SWDF Solid Waste Disposal Facility

TEPC Tissue Equivalent Proportional Counter

TBD Technical Basis Document

TLD Thermoluminescent Dosimeter

TLND Thermoluminescent Neutron Dosimeter

TPO Technical Procedures Office

TRU Transuranic

WCF Waste Certification Facility

WSRC Westinghouse Savannah River Company
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 19 of 188

EXECUTIVE SUMMARY

In enacting the Energy Employees Occupational Illness Compensation Program Act of 

2000 (EEOICPA), the U.S. Congress officially recognized the hazardous nature of 

producing and testing nuclear weapons. Under the Act workers, who have developed 

selected types of cancer or their survivors may be entitled to compensation and medical 

benefits. This program is administered by the Department of Labor (DOL) Office of 

Worker Compensation (OWCP). The Department of Health and Human Service’s (HHS) 

National Institute for Occupational Safety and Health (NIOSH) is responsible for 

determining the individual worker’s dose.

Specifically the Act requires the estimation of radiological doses from ionizing radiation 

received by workers in the nuclear weapons production programs of the various U. S. 

Department of Energy (DOE) and its predecessor agencies. Methods for implementing 

provisions of the Act have been promulgated in 42 Code of Federal Regulations Part 82 

(40CFR Part 82), “Methods for Radiation Dose Reconstruction Under the Energy 

Employees Occupational Illness Compensation Program Act of 2000” (Federal Register, 

Vol. 67 No. 85, Thursday, May 2, 2002).

Oak Ridge Associated Universities (ORAU) leads a team, the ORAU Team, to support 

NIOSH in conducting this major program. This technical basis document (TBD) 

represents a specific support mechanism to the ORAU Team concerning documentation 

of historical practices at the Savannah River Site (SRS). This TBD can be used to 

evaluate both internal and external dosimetry data for unmonitored and monitored 

workers and serve as a supplement to, or substitute for, individual monitoring data. This 

document provides a Savannah River Site (SRS) profile that contains technical basis 

information used by the ORAU Team to evaluate the total occupational radiation dose for 

Energy Employees Occupational Illness Compensation Program Act (EEOICPA) 

claimants. 

This document also provides supporting technical data to evaluate, with claimantfavorable assumptions, the total SRS occupational radiation dose that may reasonably be 

associated with the worker’s radiation exposure. This dose results from exposure to 

external and internal radiation in SRS facilities; to SRS occupationally-required diagnostic 

x-ray examinations, and to on-site environmental releases. Also included is the dose that 

may have occurred while the worker was unmonitored or the dose may have been 

missed. Over the years new and more reliable scientific methods and protection 

measures have been developed. The methods needed to account for these changes are 

also identified in this document. 

The doses are evaluated using the NIOSH Interactive RadioEpidemiological Program 

(IREP) and the Internal Modular Bioassay Analysis (IMBA) computer codes. Information 

on measurement uncertainties is an integral component of the NIOSH approach. This 

document describes how the uncertainty for SRS exposure and dose records is 

evaluated.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 20 of 188

The document is divided into five major sections; Facilities and Processes, Occupational 

Medical Dose, Occupational Environmental Dose, Occupational Internal Dose, and 

Occupational External Dosimetry. Each section has an associated attachment, which 

provides the critical data for the specialists reconstructing the doses. 

The facilities and processes used at SRS over the years and their associated source 

terms are briefly described in section one. The tables showing the radionuclides of 

concern for dose reconstruction and the years of operation of the different facilities are in 

Attachment A. The discussion includes the individual processes and the radionuclides for 

both the internal and external exposure potentials. These data would be used by the dose 

reconstructors only when monitoring data are unavailable and/or other methods may not 

be amenable for claimant dose reconstruction.

Section 2 discusses the occupational medical dose. The medical doses result from the 

chest X rays given during pre-employment and annual physical exams. However, certain 

employees may have had X rays more frequently based on their work and their age. 

Although X rays may have been given because of on-the-job injuries, these doses are not 

considered part of the occupational dose. The x-ray techniques and equipment used over 

the years have also changed and are presented and discussed in this section. 

Attachment B presents the pertinent tables for use by the dose reconstructors in 

determining the worker’s occupational medical doses.

Section 3 provides information for calculating internal and external radiation doses to 

unmonitored workers at the SRS. Data are presented on the radionuclides released from 

the site facilities and processes. The methodology to calculate the doses received by 

unmonitored workers outside the facilities is described. Internal doses result from 

breathing the radioactive material in the air, ground level releases, or resuspended from 

radioactive material deposited on the ground. External doses result from the ambient 

radiation levels and from submersion in a plume of noble gases. The equipment used to 

measure both the internal and external doses changed considerably over the years and 

these changes have been taken into account in calculating the individual worker dose.

A dispersion model was developed using a 5-year average joint frequency distribution. 

The values were converted to X/Q tables for elevated and ground level releases, using 

the Multimedia Environmental Pollutant Assessment System (MEPAS) 4.0 computer 

code, and plotted using an Excel spreadsheet. Various assumptions were made to 

ensure atmospheric dispersion values were not being overestimated. The maximum air 

concentrations were then calculated for the principal dose producing radionuclides at 

various on-site populated areas for use in calculating the unmonitored worker doses. 

Attachment C provides a series of tables of the 50th percentile and geometric standard 

deviation (GSD) intake values for selected individual radionuclides resulting from 

releases from various facilities. These values will be used by the dose reconstructors to 

calculate the internal and external doses that unmonitored workers may receive from the 

environmental pathway.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 21 of 188

Section 4 discusses the occupational internal dose. The radionuclides identified by 

Savannah River as being the most significant in delivering doses or that were effective 

tracers for significant dose-producing radionuclides are presented. The methods for 

analysis of selected individual radionuclides are also discussed. Tables providing the 

minimum detectable activities (MDAs) are included for the In vitro analysis of urine and 

feces for selected radionuclides and in vivo whole body and chest counting.

Attachment D provides a more complete list of radionuclides and MDAs for use by dose 

reconstructors for calculating internal doses. 

Section 5 presents the occupational external dosimetry program for measuring skin and 

whole body doses to the workers. The methods for evaluating external doses to workers 

have also evolved over the years as new techniques and equipment have been 

developed. Concepts in radiation protection have also changed. The dose reconstruction 

parameters, SRS practices and policies, and dosimeter types and technology for 

measuring the dose from the different types of radiation are discussed in this section. 

Attention is given to the evaluation of doses measured from exposure to beta, gamma, 

and neutron radiation. Test results for various films exposed to different geometries and 

energies of radiation are given in tables.

Sources of bias, workplace radiation field characteristics, responses of the different 

beta/gamma and neutron dosimeters in the workplace fields, and the adjustments to the 

recorded dose measured by these dosimeters during specific years are discussed in 

detail.

There are sources of potential dose that could be missed because of the limitations of 

dosimetry systems and the methods of reporting low doses. This missed dose is 

discussed as a function of facility location, dosimeter type, year, and energy range. 

Attachment E describes the use of the external dosimetry technical basis parameters to 

facilitate the efforts of the dose 

reconstructors.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 22 of 188

1.0 INTRODUCTION

The National Institute for Occupational Safety and Health (NIOSH) was assigned the 

responsibility for developing the technical capabilities and guidance to be used to 

implement the Energy Employees Occupational Illness Compensation Program Act 

(EEOICPA). Oak Ridge Associated Universities (ORAU) led a team, identified as the 

ORAU Team, to support NIOSH in the conduct of this major program. This technical 

basis document (TBD) represents a specific area of support to the ORAU Team 

concerning documentation of historical practices at the Savannah River Site (SRS). It 

pertains to evaluation of both internal and external dosimetry data for unmonitored and 

monitored workers and is to be used as a supplement to or substitute for individual 

monitoring data. 

1.1 Purpose

The purpose of this document is to provide a Savannah River Site (SRS) profile that 

contains technical basis information used by the Oak Ridge Associated Universities 

(ORAU) Team to evaluate the total occupational dose for EEOICPA claimants. 

1.2 Scope

SRS operations played an important role in the US nuclear weapons program (DOE 

1997). These SRS processes included nuclear fuel fabrication, reactor operation, 

radiochemical processing, uranium recycling, plutonium production, neutron source 

production, and waste management. Supporting documentation has been assembled in 

this TBD to assist in the evaluation of worker dose from these processes using the 

methodology in the respective NIOSH OCAS-IG-001 External Dose Reconstruction 

Implementation Guideline (NIOSH 2002a) and OCAS-IG-002 Internal Dose 

Reconstruction Implementation Guideline (NIOSH 2002b). 

The methods and concepts of measuring radiation exposure to workers have evolved 

since the beginning of SRS operations in 1952. An objective of this document is to 

provide supporting technical data to evaluate, with claimant-favorable assumptions, the 

total SRS occupational dose that may reasonably be associated with worker radiation 

exposure as covered under EEOICPA legislation. This dose includes occupational 

external and internal exposure in SRS facilities; SRS occupationally-required diagnostic 

X-ray examinations, and on-site exposure to SRS environmental releases. The 

documentation addresses evaluation of unmonitored and monitored worker exposure, 

and missed dose. Consistent with NIOSH Implementation Guidelines, this document 

identifies how to adjust the historical occupational dose to account for current scientific 

methods and protection factors. 

This document also presents the technical basis of methods used to prepare the SRS 

worker dose records for input to the NIOSH Interactive RadioEpidemiological Program 

(IREP) and the Integrated Modules for Bioassay Analysis (IMBA) computer codes used 

to evaluate worker dose. Because information on measurement uncertainties is an 

integral component of the NIOSH approach, this document describes how the 

uncertainty for SRS exposure and dose records is evaluated. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 23 of 188

The main body of text in this document provides the description of the facilities and 

processes, historical information related to worker internal and external exposures, and 

environmental data for use when actual monitoring data are unavailable. The 

attachments represent the critical data and tables required by dose reconstructors for 

performing the individual claimant dose reconstructions. These attachments should 

suffice as a stand-alone document for dose reconstruction. Additional details, if 

necessary, could be found in the main body of the text.

1.3 Site Activities and Processes 

This section briefly describes the facilities and processes while their associated source 

terms have been captured in Attachment A. These data would be used by dose 

reconstructors only when monitoring data are unavailable and/or other methods may not 

be amenable for claimant dose reconstruction. As indicated, the tables listed below and 

referenced in the text refer to the tables in Attachment A. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 24 of 188

The following tables or sections are provided in Attachment A:

Table Number Facility or Process Page Number

A-1.1 C, L, K, P and R Reactors-Years of Operation

A-1.2 Radionuclides of Concern for All Reactors

A-2 F-Area A-Line Facility

A-3 221-F B-Line Facility

A.4-1 221-F Canyon Facility, Head End Stream (Stream 1)

A-4.2 221-F Canyon Facility, (2nd Uranium Cycle) (Stream 2)

A-4.3 221-F Canyon Facility, (2nd Cycle Plutonium) (Stream 3)

A-5 New Special Recovery (NSR) Facility (Pu Recovery)

A-6.1 F-Area Outside Facilities, Acid Recovery Unit (ARU)

A-6.2 F-Area Outside Facilities, Water Handling System

A-6.3 F-Area Outside Facilities, General Purpose Evaporators

A-6.4 F-Area Outside Facilities, Segregated Plutonium Solvent

A-6.5 F-Area Outside Facilities, Laboratory Waste

A-6.6 F-Area Outside Facilities, Segregated Uranium Solvent

A-7 238PuO2 Fuel Form Facility (PuFF) and 238PuO2

Experimental Facility (PEF)

A-8 235-F Vaults (Pu Oxide, Finished Product, Scrap Vaults)

A-9 772-F and 772-1F Production Control Laboratories

A-10 E-Area Solid Waste Disposal Facility (SWDF)

A-11 Plutonium Storage Facility (PSF)

A-12 F/H Effluent Treatment Facility (ETF)

A-13 F & H Area Cooling Water and Retention Basins

A-14 F & H Area Tank Farms

A-15 Waste Certification Facility (WCF)

A-16.1 221-H B-Line Facility, Scrap Recovery U-238, U-239

A-16.2 221-H B-Line Facility, Scrap Recovery U-235

A-16.3 221-H B-Line Facility, NpO2 and PuO2 Facility

A-17.1 H-Canyon Facility, First Cycle Stream-Uranium Cycle

A-17.2 H-Canyon Facility, Second Uranium Cycle Stream

A-17.3 H-Canyon Facility, Neptunium Second Cycle

A-17.4 H-Canyon Facility, Work Stream

A-18.1 H-Area Outside Facilities, A Line

A-18.2 H-Area Outside Facilities, Water Handling System

A-19.1 Receiving Basin for Offsite Fuel (RBOF)

A-19.2 Resin Regeneration Facility (RRF)

A-20 Tritium Facilities

A-21 Uranium Target Fabrication Facility, 313 M

A-22 Uranium Processing Facilities 320-M, 322-M, and 341-M

A-23 Fuel Fabrication Facility, 321-M

A-24.1 S-Area Defense Waste Processing Facility (DWPF), 

Sludge Slurry Feed Stream (Stream 1)

A-24.2 S-Area Defense Waste Processing Facility (DWPF), 

Precipitate Slurry Feed Stream (Stream 201)

A-25.1 Z Area, Waste Tank Supernate - Decontaminated Salt 

Solution

A-25.2 Z Area, Saltstone

A-26 A Area Building 773-A

A-27 Buildings 735-A and 735-11A

A-28 Building 776-A Liquid Waste Handling Facility

A-29 D-Area, Heavy Water Production and Reprocessing
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 25 of 188

The data in the tables in Attachment A have all been extracted from ESH-HPT-960197, 

Facility Descriptions (U), Thomas R. La Bone, October 1, 1996.

All stated assumptions are also from La Bone (1996).

The following definitions apply to all tables in Attachment A that contain information on 

the Activity Fraction:

The activity fraction is the fraction of the total activity represented by a particular 

radionuclide.

The units in the tables in Attachment A are conventional units. To convert from 

mrem/nCi to Sv/Bq, multiply by 2.7E-7. 

To convert from nCi to Bq, multiply by 37.

Where solubility classes were available in the Savannah River Site literature, their 

designations have been changed to the ICRP absorption type designations in the tables. 

According to the International Commission on Radiological Protection ICRP Publication 

68 “Dose Coefficients for Intakes of Radionuclides by Workers” the ICRP 30 D, W, and 

Y absorption classes correspond broadly to the absorption Types F, M, S respectively. 

These values are given in the second column of the tables in Attachment A.

The differences in the radionuclides for the various processes in Attachment A are 

related to the mode of exposure. To minimize the number of pages that the dose 

reconstructor would need to search, both internal and external radionuclides have been 

combined into one table for each of the facilities/processes addressed for the SRS. 

Radionuclides identified as “being of concern” for internal exposure may be different 

from those identified as providing the greatest external exposure. The last column of 

some of the tables is “Significant to External Exposure.” An “X” in this column indicates 

that the radionuclide contributes to external dose. While 90Sr is typically an internal 

exposure hazard, under certain circumstances it could also be an external exposure 

hazard. It is listed in this document as both for completeness. 

Dates of operation of the various facilities have been taken from:

Cummins, et al, Radioactive Releases from the Savannah River Site, 1954 – 1989, 

WSRC-RP-91-684,

DOE/EM-0319, Linking Legacies, January, 1997, and

Till, et al, Savannah River Site Environmental Dose Reconstruction Project, Phase II, 

Risk Assessment Corporation, 30 April 2001.

The discussions that follow relate directly to the table represented in the individual 

section heading, e.g., A-1 Reactors will relate to Table A-1 in Attachment A, etc. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 26 of 188

A-1 Reactors (LaBone 1996, Till 1999)

Five heavy-water reactors designated R, P, L, K, and C were constructed at SRS in the 

early 1950s, each of which is in the locations designated R-Area, P-Area, L-Area, KArea, and C-Area respectively. All five reactors were operating in 1955. Reactor 

shutdown began in 1964 with R-Reactor. C-Reactor was shut down in 1987, L-Reactor 

was shut down in 1968, restarted in 1985, and then shut down again in 1988. K-Reactor 

and P-Reactor operated with few interruptions until 1988. A restart of K-Reactor began 

in 1991, but was not completed. None of the reactors are currently operational.

Depleted uranium, 237Np, and lithium targets were irradiated to produce 239Pu, 238Pu, and 

tritium, respectively. The targets are irradiated and transferred to the Tritium Facilities, H 

Area Separations (238Pu), and F Area Separations (239Pu) for processing. Spent reactor 

fuel is sent to H Area Separations for processing. In the past the reactors were also used 

to convert thorium to 233U as well as to irradiate targets that produced transplutonic 

elements such as curium and californium.

The composition and relative abundance of the radionuclides in the reactor environment 

are not known. Representative samples are taken annually to determine the isotopic 

makeup of contamination in reactor areas. When the reactors are operating, fission 

neutrons are an external exposure concern in some areas. For the purpose of 

determining internal or external exposure in the reactor areas, the radionuclides of 

concern in the reactor areas are presented in Table A.1-1.

Internal exposure

Because the reactors were heavy-water reactors, intakes from the reactor buildings 

would be so dominated by tritium that intakes of fission/activation products can be 

ignored. If, in a specific case, the bioassay data indicate otherwise, then see the fission 

product discussion for the 221-F Canyon Facility. 

 - Discussions by T.R. La Bone with Health Physics Operations 

personnel and a review of ANSI 343 were used to identify the radionuclides of concern. 

Tritium is assumed to produce most of the personnel exposure because of the large 

quantities present.

External exposure - During operation, fission neutrons are also a concern in certain 

areas. 

A-2 F-Area and A-Line Facility

The A-Line facility accepts dilute, depleted uranyl nitrate solution from the 221-F Facility 

and converts it into uranium trioxide powder. Dilute uranyl nitrate solution is sent to ALine from the solvent extraction process in the 221-F Facility where it is initially decanted 

to remove tributyl phosphate n-paraffin solvent. It is then concentrated by evaporation. 

A second evaporation yields a hydrate of uranyl nitrate that is heated in a denitrator to 

yield the uranium trioxide product.

The A-Line facility also has a dissolver to prepare a uranyl nitrate solution from the 

uranium trioxide product and nitric acid when it is required as a process feed stream 

back into the solvent extraction process in the 221-F Facility. The dissolver is also used 

to recycle off-standard A-Line product. The uranium trioxide product is placed into drums 

for storage on-site or into packages for off-site shipment.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 27 of 188

The A-Line Facility can be broken down into unit operations that perform specific 

functions in the creation of the final product. These processes consist of:

• evaporation, that accepts uranyl nitrate from the 221-F Warm Canyon and 

concentrates it; 

• purification, which is an optional step, but by using silica gel absorbs residual 

zirconium and niobium out of the uranyl nitrate solution;

• hydrate evaporation, where uranyl nitrate from the silica gel or evaporator is again 

concentrated; 

• denitration, where material from the hydrate evaporation is received and converted 

from uranyl nitrate hexahydrate to uranium trioxide; 

• material handling, where the uranium trioxide powder is removed from the denitrators 

and placed into storage bins from which it is loaded into containers, and 

• oxide dissolution, where 221-F occasionally requires a uranyl nitrate-nitric acid 

solution for use as a process stream.

• The fume recovery system converts oxides of nitrogen into nitric acid for reuse using 

exhausters, a venturi scrubber, off-gas coolers, an absorption column, and various 

pumps and tanks. (The venturi scrubber removes uranium trioxide particles from the 

denitrator off-gas and is the only piece of the system that is of radiological concern.)

Internal exposure - The chemical forms that are produced in A-Line are uranyl nitrate 

and uranium trioxide. These materials are Types F and M, respectively. See also the 

discussion in 4.1.2, “In Vitro Bioassay for Uranium,” for guidance on interpreting uranium 

urinalyses for natural or depleted uranium. Although the trace contaminants in the 

uranium provide little to “compliance” dose (i.e., 50-year committed doses), they may 

contribute dose to different organs than the uranium and they will be more significant if 

the time from intake to diagnosis is short relative to 50 years. Hence, the trace 

radionuclides should be included in intake calculations in the ratio relative to 238U as 

shown in Table A.2. 

External exposure - Depending on the age of the material in the A-Line, it is assumed a 

mixture of long-lived fission products would be the major contributors of external dose. 

The long-lived fission products of interest would include 95Nb, 103Ru, 106Ru, 141Ce, 144Ce, 

144Pr, 151Sm, and 155Eu. Other contributors would be 235U, 236U, 238U, and 239Pu.

The activity fractions given in the table indicate the predominant radionuclides of 

concern for external dose would be 238U, 141Ce, 144Ce, 103Ru, 106Ru, and 95Nb. 

A-3 221-F B-Line Facility

The F B-Line Facility is located in the 200-F Separations Area and is used to convert 

plutonium nitrate into plutonium metal or plutonium oxide. It also has the capability to 

recover plutonium from on-site and off-site scrap. Initially the plutonium produced in this 

facility is primarily 239Pu in a dilute nitric acid solution. From there it undergoes many 

chemical processes that produce an end product.

The individual processes in the 221-F B-Line Facility consist of:

• Cation exchange, where dilute plutonium solution transferred from the canyon 

storage tanks to the F B-Line receiving tanks is sent through a cation exchange 

column that removes the plutonium along with some impurities; 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 28 of 188

• Drying and conversion, where the plutonium trifluoride taken from the filtration 

process is dried and roasted to produce plutonium dioxide and plutonium 

tetrafluoride;

• Reduction, where the plutonium dioxide-plutonium tetrafluoride mixture is weighed, 

mixed with metallic calcium and placed into a magnesium crucible whose inside is 

filled with magnesium oxide sand, sealed, pressurized with argon, and heated to a 

molten state, Pu metal flows to the bottom of the crucible and forms a button shaped 

solid, which is then cooled, separated from the calcium slag, and die-stamped for 

identification;

• Button finishing, where the stamped button is submerged in nitric acid, rinsed in 

water and dried (after which one or two grams of the plutonium button are removed 

using a drill), placed in a can, crimp-sealed, placed in a plastic bag, a second can, 

crimp-sealed, and sent to storage.

The drill turnings are sent to the 772-F Laboratory for analysis and recovery, where 

plutonium is extracted from the solid scrap from on-site and off-site sources, as well as 

miscellaneous solutions from F B-Line by using an anion exchange resin. The scrap is 

initially dissolved using a mixture of aluminum nitrate and nitric acid, after which the 

plutonium scrap is sorbed on the anionic resin and sent back to the canyon for 

processing.

External exposure considerations - The alpha emitting isotopes of plutonium are not 

usually an external exposure concern. However, when they exist in a fluoride form, the 

alpha particles interact with the fluorine by an (α, n) reaction to produce a significant 

source of neutrons. Therefore, neutrons are a major external exposure concern in the F 

B-Line.

Neutrons are also emitted by spontaneous fission along with associated beta and 

gamma radiations from 238Pu and 240Pu and progeny. Plutonium-241 decays by beta 

emissions to 241Am which emits a 60 keV photon and X rays with an energy of 17 keV. 

These photons are the primary external exposure concerns for 241Pu.

Many different mixtures of plutonium isotopes are possible in the 221-F B-Line, but it is 

assumed that the mixture will be either 6% or 12% Pu. The percent refers to the mass 

percent of 240Pu. The radionuclides of concern in Table A-3 are the same for both 

internal and external exposure.

Internal exposure - the plutonium chemical compounds in the process are varied, and 

the lung inhalation/solubility types vary with these compounds so Table A-3 does not list 

a specific type. If the bioassay data are not robust enough to imply the type, the dose 

reconstructor should assume type M for nitrate form and type S for oxide form. 

However, old contamination, even if originally in the nitrate form, slowly oxides and could 

be considered type S. If no information is available concerning the chemical form, 

assume a type that is claimant-favorable; type M will increase the dose to systemic 

organs; type S will increase the dose to the lung and GI track organs. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 29 of 188

A-4 221-F Canyon facility

The 221-F Canyon Facility consists of two parallel canyons that constitute the process 

area. These canyons contain the high-activity and low-activity radioactive materials and 

are referred to as the hot and warm canyons, respectively.

The F-Canyon facility uses the Plutonium Uranium Extraction (PUREX) process to 

recover 239Pu, 237Np, and 238U from reactor-irradiated targets. The plutonium is 

transferred to the B-Line and the uranium is transferred to the A-Line facility where they 

are processed into a solid form.

The 221-F Canyon processes are:

Head End 

Irradiated targets are received from the reactor areas. The aluminum cladding is 

dissolved in hydroxide-sodium nitrate solution, and the uranium metal is then 

dissolved in nitric acid.

Solvent Extraction

The first solvent extraction process separates the plutonium and the uranium 

from the fission products using tributyl phosphate in a hydrocarbon diluant. In the 

second cycle, plutonium and uranium are separated and purified.

Anion Exchange

The aqueous waste produced in the first solvent extraction cycle and the second 

uranium cycle contain neptunium and plutonium in a nitric acid solution, which 

are recovered using an anion exchange resin column. After the solution is sent 

through the column, the diluted actinides are sent to the first solvent extraction 

cycle or warm canyon for further processing and separation. If the concentration 

of the actinides is such that the solution is sent to the warm canyon, a further 

series of columns are used to purify and separate the plutonium from the 

neptunium. At this stage, 232Th and 234Th are also removed.

Waste Evaporators

Aqueous waste from the first extraction cycle, the second uranium cycle, acid 

strip condensate, ventilation system sumps, and concentrate from the second 

stage high activity waste evaporator are all combined and evaporated in the first 

stage continuous evaporator. This process reduces the volumes of waste going 

to the high level waste storage tanks.

Locations of Increased Exposure Potential

The hot and warm crane cabs in 221-F have increased potential for exposure to 

radioactive materials. These areas are the major concern because the majority of 

the process is isolated and run remotely due to the high radiation fields.

Internal exposure - The majority of the dose from an intake of material from the Head End 

Stream comes from plutonium, cerium, and ruthenium, but because different radionuclides 

target different organs in the body, none of the radionuclides listed in Table A-4 should be 

ignored. Intakes of the fission and activation products would be detectable with good 

sensitivity by whole body counts, except for 90Sr and 91Y (the progeny of 91Sr). Not listed, 

but undoubtedly present, was 89Sr. However, 89Sr would have been counted as part of the 

gross beta count for 90Sr, and assigning a radiostrontium gross beta result to 90Sr is claimant 

favorable. If fission products were detected by whole body counts and no urinalysis for 90Sr 

was available, an intake of 90Sr equal to the intake of 137Cs should be made. Because 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 30 of 188

137Cs is generally more mobile in the reactor and Head End work environment, this 

assumption should be conservative. This assumption would account for the dose from 

91Sr/91Y as well, which because of the short physical half-lives, produce negligible dose 

compared to the 90Sr. 

Some fission products will be more volatile than others during the dissolution so hopefully 

whole body counts will provide the ratios of the gamma-emitting radionuclides in the body. If 

fission products are detected in a worker from this facility, and no 90Sr or plutonium 

bioassay data are available, assume the ratios provided in Table A-4. An intake of 241Pu 

using the ratio to 239Pu provided in Table A-3 or for standard 6% or 12% plutonium mixture 

(Table 4.1.1-4) should also be included. To be claimant-favorable, an intake of 91Y the 

same magnitude as 106Ru should be assumed. 

In the second uranium cycle stream (Stream 2) the majority of the dose comes from 238U 

and 239Pu; however, the same considerations concerning fission products and 241Pu stated 

in the preceding paragraph should be applied. 

The third stream is known as the second cycle plutonium stream (Stream 3) in which 

plutonium dominates.

Personnel in this facility are exposed to all processes. Since that is the case, it is not 

possible to determine the uranium ratios required to determine the Minimum Detectable 

Intake (MDI). Since 238U is the main isotope in every process stream, the MDI was 

figured specifically for 238U. 

External exposure - Fission products are present in the 221-F Canyon Facility and, 

depending on the age of the material (i.e., the length of time out of the reactor), the 

isotopes of concern and the fraction of the activity of these materials will change. 

The majority of the external dose from Stream 1 comes from cerium, zirconium, and 

ruthenium, all photon and beta emitters. In the second stream 238U, which feeds 234mPa, 

makes up the majority of the external dose along with zirconium which emits both 

photons and betas. In the third stream the majority of the external dose comes from 

241Pu and 95Zr.

A-5 New Special Recovery (NSR) Facility

The NSR facility recovers plutonium from a variety of sources including scrap plutonium 

from offsite sources and from the F B-Line facility. The processed plutonium is 

conditioned so that it can be blended back into the plutonium nitrate feed stream to the F 

B-Line process.

The New Special Recovery Facility consists of the following processes: 

• Material Receiving and Assaying, where the material is assayed to determine its 

plutonium content before any processing begins; 

• Feed Preparation, which is dependent on the type of scrap; 

• Dissolution and Filtration, which converts the solid feed material into a solution that is 

suitable for further purification by either anion exchange or by PUREX solvent 

extraction; 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 31 of 188

• Anion Exchange Purification, which removes impurities such as fluoride, aluminum, 

americium, sulfate, and uranium;

• Waste Handling, which comes from sludge, residues, slurries, and from solid 

materials such as tools, filters, gloves, etc., and 

• Analytical and Assay Operations, where samples are analyzed and returned to the 

process.

The function of the NSR is to recover plutonium from a variety of materials. These 

materials also contain impurities such as americium and uranium. The isotopic content is 

not constant in the material or between materials. For example, the unirradiated reactor 

cores percent weight range from 6% to 12% for 240Pu, 0% to 10% for 241Am, and 60% to 

80% for uranium. It is assumed that the feed material is principally enriched uranium.

Internal exposure - The isotopes that deliver the majority of the internal dose are 239Pu, 

240Pu, 241Am, 235U, and 236U (assuming enriched uranium). 

External exposure - The isotope that delivers the majority of the photon (external) dose 

is 241Am. Some neutron dose will also be present due to spontaneous fission and an (α, 

n) reaction with fluoride, if present. 

A-6 F-Area Outside Facilities

The F-Area Outside Facilities provide general support, principally to the processing of 

irradiated fuels and targets in Building 221-F. The term “Outside Facilities” is used to 

describe a wide variety of processes, utilities, and services that are ancillary to the 

primary 200-F Area operations. The main processes of radiological concern are as 

follows: 

Water Handling Facilities

This facility is used to provide process water and acidified water streams for the 

canyons, retain discard water in holding tanks for analysis, and decant spent 

solvent from waste water.

Acid Recovery Unit (ARU)

This facility concentrates nitric acid for reuse. All material in this location can 

contain radioactive material. 

General Purpose Evaporator:

The General Purpose (GP) evaporators concentrate aqueous waste having 

activity that is in excess of disposal limits yet is low enough to be evaporated in 

unshielded equipment.

General Purpose Waste Tanks

The tanks in this facility are used to collect GP evaporator feed and to receive 

and dispense evaporator overheads for recycle to the process.

Waste Handling Facilities

The F-Area Waste Handling facilities are used for storage and transfer of highand low-activity wastes that are primarily from the Metallurgical Building 235-F, 

Savannah River Laboratory (SRL) Waste Concentration Building 776-A, and the 

Production Control Facilities, Building 772-F and 100-Area sources.

Segregated Solvent Facilities

Segregated Solvent facilities provide solvent purification and tank storage before 

the solvent is returned to F-Canyon for reuse.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 32 of 188

Transfer Tanks

These tanks provide intermediate pumping stations for those solutions 

transferring from Building 221-F.

Sump Collection Tanks

Two sump collection tanks are located in separate underground vaults. One tank 

receives radioactive waste condensate, by jet stream, from sumps or tanks 

related to the canyon air exhaust system and transfers the waste to the hot or 

warm canyon waste evaporator feed tanks. Discharges to hot canyon cell sumps 

are first transferred to the second sump collection tank and then to the canyon for 

analysis and disposition.

Recycle sump

The Recycle Sump collects drainage and overflow from the canyon auxiliary 

tanks and holds contaminated or recycled liquids.

Auxiliary Systems and Support Facilities

The only auxiliary system and support facility of concern is the plant laundry 

located in Building 723-F.

Internal exposure - The radionuclides of concern for the ARU, water handling, and the 

basin transfer tanks are 95Zr, 95Nb, 103Ru and 106Ru. 

External exposure - The radionuclides of concern for the ARU, water handling, and the 

basin transfer tanks are 95Zr, 95Nb, 103Ru, and 106Ru. The primary radionuclides of 

external dosimetric concern for the GP evaporators, Segregated Plutonium Solvent, and 

Laboratory Waste are 241Pu, 103Ru, 106Ru, 95Zr, and 95Nb. The radionuclides of concern 

for the Segregated Uranium Solvent are 103Ru, 106Ru, 95Zr, 95Nb, and 238U which feeds 

234Pa.

The radionuclide distribution in the basin transfer tanks is assumed to be similar to the 

water handling distribution.

A-7 238PuO2 Fuel Form Facility (PuFF) and the 238PuO2 Experimental Facility (PEF) 

The primary function of the PuFF facility is to produce encapsulated plutonium oxide fuel 

forms. This is accomplished by converting 238Pu oxide powder into fuel forms for heat 

sources by powder ceramic and metallurgical process. The final products are compacted 

PuO2 fuel shapes. The PEF provides space for the testing of processes to be used in the 

PuFF facility. The plutonium oxide blends used in the Puff and PEF facilities are very 

similar.

The PuO2 is fabricated and encapsulated in six manipulator cells filled with inert gas, 

three air-filled manipulator cells, five wing cabinets, and one hood.

The fuel pellets are made by hot-pressing a blended PuO2 shard mixture prepared from 

calcined Pu oxalate powder received from H B-Line. Feed processing includes oxygen16 enrichment, ball milling, compaction, granulation, and sintering. After final heat 

treatment, the fuel pellets are encapsulated in iridium-clad vent sets by tungsten inert 

gas welding. The conversion of plutonium oxide powder into dense PuO2 fuel forms is 

conducted in the PEF.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 33 of 188

Internal exposure - The majority of the dose from the PuFF and PEF material comes 

from plutonium. 

External exposure - The majority of the external dose from the PuFF and the PEF comes 

from the neutrons generated by spontaneous fission of 238Pu and the photon/beta 

emissions from 241Am, 232Th, and 232Np. 

A-8 235-F Vaults

Vaults in the 235-F Facility were built to store plutonium at different production steps 

prior to completion and were created not only for safety reasons but also to establish 

safeguards for security purposes. The 235-F facility receives plutonium and plutonium 

dioxide and fabricates it into a variety of items used in the nuclear program. There are 

three vaults located in the 235-F facility:

The Plutonium Oxide Vault

Used to store plutonium oxide powder from the PuFF and PEF, and the Actinide 

Billet Fabrication Facility. Neptunium-237 powder and billets are also stored in 

this facility.

The Finished Product Vault

Used for the storage of uranium and transuranics in powder form like the 

plutonium oxide vault. In addition, this vault contains a water cooled, shielded 

storage area for the storage of finished 238Pu products.

The Scrap Vault

Used to store metal and oxide scraps of uranium, thorium, and plutonium that are 

in the form of chips, sweepings, powder, blended samples, or metal turnings.

Internal exposure - Radioisotopes that are stored and found in the 235-F vaults are 

238Pu, 239Pu, HEU, and 237Np. All are of concern for determining internal dose.

External exposure - Detailed information on the composition of the radionuclides stored 

and found in these vaults is not available; therefore, the radionuclides listed in the 

associated table in the Attachment are considered to be pure. The major source of 

external exposure is 238Pu spontaneous fission neutrons and betas from 238Pu progeny in 

addition to gammas and X rays.

A-9 772-F and 772-1F Production Control Laboratories

The production control laboratories provide analytical support for the 200-F and 200-H 

separations processes. Support is also provided to Waste Management, Reactors, and 

Raw Materials.

The radionuclides of concern are plutonium, uranium, and fission products. The major 

source of external exposure is 238Pu spontaneous fission neutrons and betas from 238Pu 

progeny in addition to gamma, X ray photons, and betas from fission products.

A-10 E-Area Solid Waste Disposal Facility (SWDF)

Solid and liquid low-level waste, generated at the SRS and the SRL, and some waste 

from off-site are stored at the SWDF.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 34 of 188

Solid wastes are located above and below ground. The SWDF is located in areas 

labeled 643-E and 643-7E. The old burial ground is designated 643-E and ceased 

operations in 1972. Burial operations were phased in during 1969 to 1972 in the present 

SWDF designated as 643-7E. Most of the solid waste above ground is TRU waste, 

mainly 238Pu and 239Pu contaminated waste. Contaminated process equipment contained 

in steel boxes is stored above ground.

Liquid waste is stored in below ground tanks. Thirty two tanks have been used to store 

liquids. As of 1988, only three tanks contained liquids. Liquid radioactive waste is in the 

form of degraded organic process solvents, waste oil, scintillation solutions, and 

contaminated hazardous waste chemicals.

Due to the diverse nature of the radioactive waste, the radionuclide content of each type 

of waste has not been defined. The radionuclides of highest concentration (Ci/vol) are 

tritium, cobalt, plutonium, and curium. 

External Exposure - The radionuclides of external dosimetric significance are the fission 

and activation products associated with reactor operations, such as 60Co, Cs, Sr, Ni, Ce, 

Rh, and Zr.

A-11 Plutonium Storage Facility (PSF)

The PSF is located in the 200-F Separations Area and is used to receive, store, monitor, 

retrieve, and ship packaged plutonium. Plutonium in solid and powdered form (such as 

scrap) contained in approved shipping containers may be temporarily stored there prior 

to shipment.

The Receipts Assay Facility is the primary location for material accountability. 

Specifically, it is the area used for the unpacking of off-site material. Material from offsite is unpacked in a shielded glovebox, the inner package is removed, and it is passed 

on to the loading station where the inner package is weighed, labeled and sealed, then 

sent to the nondestructive assay laboratory.

Internal exposure - The radionuclide of concern in PSF is plutonium. Many different 

mixtures of plutonium isotopes are possible, but it is assumed that the mixture is either 

6% Pu or 12% Pu, the percent referring to the mass percent of 240Pu.

External exposure - The alpha-emitting isotopes of plutonium are not usually an external 

exposure concern. However, when they exist in a fluoride form, the alpha particles 

interact with the fluorine by an (α, n) reaction to produce a significant source of neutrons. 

Neutrons are also emitted by spontaneous fission along with associated beta and 

gamma radiations from 238Pu and 240Pu.

Plutonium-241 decays by beta emissions to 241Am which emits a 60 keV photon and X 

rays with an energy of 17 keV. These photons are the primary external exposure 

concerns for 241Pu.

All radionuclides identified in Table A-11 are applicable for determining internal and 

external dose.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 35 of 188

A-12 F/H Effluent Treatment Facility (ETF)

The ETF consists of large, impermeable storage basins (with lift stations located in both 

F and H Areas) and a treatment facility (Building 241-84H located in H Area). ETF is 

designed to remove hazardous chemical and radioactive contaminants from the 200-

Area liquid effluent waste streams and concentrate them for disposal. The effluent 

processing steps include feed adjustment (which consists of oxidation, grit collection and 

removal) and pH adjustment; filtration, which consists of three parallel trains with three 

stages of filtration; organic removal, consisting of mercury removal ion exchange 

columns (3) in parallel followed by 2 activated carbon beds in series; reverse osmosis, 

which removes the bulk of the solid material in the 200-Area liquid effluent wastes using 

a three-stage reverse osmosis process that concentrates dissolved solids; ion exchange, 

which removes ionic species of radionuclides and mercury; and evaporation (which is 

fed concentrate from the filtration and reverse osmosis systems, regeneration solution 

from the ion exchange process, and the cleaning solution from the filtration and the 

reverse osmosis processes). Evaporator steam condensate is collected in a condensate 

receiver and pumped to the H-Area segregated water system for analysis. The 

overheads are transferred to the organic removal feed tank or recycled to the waste 

water collection tanks for reprocessing. The evaporator bottoms are sent to the H-Area 

waste tank farm for transfer to the Saltstone Facility. 

Internal exposure - The majority of the annual internal dose from the F/H effluent 

material comes from plutonium, cerium, and ruthenium. 

External exposure - The majority of the external dose from the F/H effluent material 

comes from cerium, cesium, zirconium, strontium, and ruthenium.

A-13 F/H Cooling Water and Retention Basins

The F and H Areas consist of chemical separation plants and appropriate support 

facilities for the production of purified radionuclides, principally uranium, plutonium, and 

tritium. The cooling water and retention basins are support facilities for the separations 

and waste management areas. Low-level radioactive liquid wastes from the separations 

areas are routed to the basins when activity is detected.

Cooling Water Basins

When radioactivity is encountered in a cooling water system, immediate action is 

taken to divert the water and isolate the leak. The cooling water system is flushed 

until the release guidelines are met.

Retention Basins

Both F and H Areas use a retention basin when it is necessary for temporary 

storage of storm drains that may be contaminated. After sampling and analysis, 

the water may be further processed, transferred to ETF, or released.

Internal exposure - The majority of the annual internal dose from the F and H Area 

basins comes from plutonium and uranium. 

External exposure - The majority of the external dose from the F Area and H Area basins 

comes from 137Cs, 90Sr, 106Ru, 144Ce, and 238U feeding 234mPa. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 36 of 188

A-14 F/H Area Tank Farms

High-level liquid radioactive wastes from SRS are received and managed in large 

underground tanks in the waste tank farms in 241-F and 241-H. The two farms contain 

51 large subsurface tanks and related facilities required for safe handling, processing, 

and temporary retention of liquid wastes. The liquid waste comes primarily from fuel 

reprocessing operations in F and H Areas with quantities from the Receiving Basin for 

Off-site Fuel and Resin Regeneration Facility (RBOF-RRF) operations, SRL, and the 

100 Areas. Liquid wastes are classified as either high heat waste (HHW) or low heat 

waste (LHW), depending on whether or not the concentration of radioactivity is such that 

forced cooling is required to maintain waste temperature within operating guidelines.

F Area and H Area Reprocessing Waste

Liquid waste from Buildings 221-F and 221-H is made up of many waste streams 

generated during the recovery and purification of the reprocessing operation, i.e., 

the HHW from 221-F & H and the LHW from F Area and H Area.

RBOF-RRF

Waste from RBOF (244-H) and RRF (245-H) is produced during regeneration of 

ion-exchange beds, backwashing filters, cleaning and handling of fuel and target 

elements, and similar incidental operations. Radioactivity in these liquids is 

approximately 400 μCi/gal, but the volume ranges from 1.0 to 2.3 million gal/yr.

SRL and 100-Area Waste

Batches of liquid waste from SRL and the reactor areas that are not suitable for 

discharge are delivered to 211-F where the waste is evaporated, adjusted for 

alkalinity, and transferred to the waste storage tanks. Reactor waste is handled 

differently if it contains insoluble materials, such as sludges or slurries, at which 

time it would be unloaded directly to the appropriate tank.

Waste Processing operations

Liquid waste from separations areas and other locations are routinely processed 

through a series of operations allowing segregation and consolidation of the 

waste components prior to interim storage.

Waste Evaporation

Radioactive waste received in the waste tank facilities is reduced by evaporation, 

and the concentrated solutions are immobilized by solidification of the residual 

salts. The evaporator condensate is continuously monitored; and if the 

radioactivity exceeds1500 dpm/mL gamma, the condensate is diverted and 

recycled.

The radionuclide contents of the waste depend on the age of the waste and vary with the 

source stream and prior processing. 

Internal exposure - The majority of the annual internal effective dose equivalent in the F 

Area combined waste tank is delivered by 90Sr, 144Ce, and 244Cm. The majority of the 

annual internal effective dose equivalent in the H Area combined waste tank is delivered 

by 90Sr, 144Ce, and 238Pu. 

External exposure - The majority of the external dose in the F Area Combined Tank 

Waste is delivered by 90Sr, 144Ce, 137Cs, and 106Ru. The majority of the external dose in 

the H Area Combined Tank Waste is delivered by 90Sr, 144Ce, and 238Pu.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 37 of 188

A-15 Waste Certification Facility (WCF)

The Waste Certification Facility (WCF), Building 724-8E, located in the 643-E Solid 

Waste Disposal Facility, is used to certify and package drums of transuranic (TRU) 

waste. The functions of the WCF are to assay and certify the contents of TRU waste 

drums and to package, load, and ship the drums. Only physical handling of sealed 

drums is performed in the WCF.

Typical generic contents of the TRU waste drums include:

Combustibles: paper, gloves, sweepings, cloth rags

Noncombustibles: HEPA filters, crucibles, glassware

Chemicals: evaporated sludge, caustic soda, soda lime, spent resins

Equipment: glove boxes, motors, scales.

Normally the drums contain either 238Pu or 239Pu depending on the source of the TRU 

waste.

Internal exposure - The radionuclides of concern for internal exposure are the same as 

for the Solid Waste Disposal facility, 643-E, and also include tritium, curium, plutonium, 

and activation products.

External exposure - The radionuclides of concern for external dosimetric purposes are 

the same as those for the Solid Waste Disposal facility, 643-E and also include curium, 

fission products, and activation products. 

A-16 221-H B-Line Facility

The new HB Line was designed to replace the aging HB-Line production facility. The HB 

Line consists of three separate facilities: the Scrap Recovery Facility, Neptunium Oxide 

Facility, and Plutonium Oxide Facility.

The Scrap Recovery Facility is designed to routinely generate nitrate solutions of 238Pu 

or 235U/239Pu scrap suitable for purification by anion exchange or solvent extraction in the 

canyon. Scrap is received and dissolver batches prepared, based on assay of the scrap. 

The solid scrap is dissolved in hot nitric acid containing trace fluoride ion, transferred 

through a filter bag, collected in a tank, sampled for accountability and process control, 

then diluted with nitric acid and transferred to the proper canyon vessel as a nitrate 

solution.

The radionuclides of concern in the scrap material are composed of either 238Pu, in oxide 

form, or 235U/239Pu mixture. The 235U/239Pu mixture normally contains greater than 20% 

enriched uranium.

In the Neptunium Oxide Facility, neptunium nitrate solutions are received from H Canyon 

and F Canyon and transferred to a receipt tank located in the HB Line. The solutions are 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 38 of 188

precipitated to yield neptunium and plutonium oxalate. The neptunium and plutonium 

oxalate cake is then loaded into a furnace and calcined to NpO2 and PuO2. 

In the H B-Line it is assumed that the NpO2 is 100% 237Np for purposes of dose 

calculations.

In the Plutonium Oxide Facility plutonium nitrate solution from H Canyon is converted to 

plutonium oxide powder by the oxalate precipitation and calcinations method. 

A-17 H-Canyon Facility

The H-Canyon building was designed for the separation and recovery of 239Pu and 238U 

from irradiated natural uranium by the PUREX process. Over the years, other processes 

have been added. These operations include the processing of irradiated enriched 

uranium to recover 235U, the processing of irradiated neptunium targets to separate and 

recover 238Pu and 237Np, and the processing of thorium to recover 233U.

In the uranium cycle, standard enriched uranium fuel is dissolved in nitric acid catalyzed 

by mercuric nitrate. The dissolved fuel is clarified in head end, and the adjusted solution 

is fed into the solvent extraction system. Uranium and neptunium are extracted from the 

fission products and separated from each other in the first cycle. The uranium product 

solution is transferred out of the building for further processing. The neptunium product 

solution is transferred to the HB Line for oxide conversion.

In the neptunium cycle, irradiated neptunium targets are dissolved in nitric acid. 

Plutonium-238 and neptunium are separated from fission products by a series of anion 

exchange resin columns. The plutonium and neptunium product solutions are then 

concentrated, precipitated to oxalate, and calcined to oxides. The plutonium oxide is 

either shipped offsite or sent to Building 235-F for formation into heat sources. The 

neptunium oxide is sent to Building 235-F for refabrication into billets.

The radionuclides of concern will vary with the age of the material.

Internal exposure - The majority of the annual internal dose for the First Cycle Stream 

comes from plutonium and cerium. The majority of the annual dose for the Second 

Uranium Cycle Stream comes from uranium. The majority of the annual dose for the 

Second Neptunium Cycle Stream comes from plutonium and cerium. The majority of the 

annual dose for the Work Stream comes from plutonium and cerium. 

External exposure - In all streams (i.e., Uranium, Neptunium, and Work Streams), the 

radionuclides of greatest concern for external exposure are the fission products cesium, 

cerium, zirconium, ruthenium, and strontium. 

A-18 211-H Area Outside Facilities

H-Area Outside Facilities include a number of processes that support the separations 

function of the 200-H Area. The main processes and services of radiological concern are 

A-Line, Water Handling Facilities, Acid Recovery Unit (ARU), General Purpose 

Evaporator, Segregated Solvent Facilities, Transfer Tanks, Sump Collection Tanks, 

Recycle Sumps, Auxiliary Systems, and Support Facilities.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 39 of 188

Many of the processes and services do not have any connection to each other in terms 

of function in the Outside Facilities. Due to this, it will in some cases be necessary to 

treat each process separately in terms of radiological protection.

A-Line

The H-Area A-Line receives a dilute aqueous solution of uranyl nitrate 

hexahydrate (UNH) enriched in 235U from the canyon where it is temporarily 

stored and then shipped off-site in tank trucks.

Water Handling Facilities

The primary equipment for water handling consists of tanks, skimmers, and 

coolers, all of which can contain radioactive materials. The facility is operated to 

provide process water and recycle water for the canyons, retain recycle water 

prior to analysis to permit disposal, decant and discharge skimmed solvent, and 

store weak acid feed for ARU.

Acid Recovery Unit

Nitric acid is concentrated in the ARU for reuse.

General Purpose Evaporator

Low-activity aqueous waste is concentrated in the General Purpose (GP) 

Evaporator reducing the volume prior to sending the waste to the tank farm.

Segregated Solvent Facilities

The final purification and storage of the solvent occurs in the Segregated Solvent 

Facilities prior to its return to the Building 221 canyons for reuse. Activity in the 

spent solvent is caused primarily by degradation of n-paraffin and Tri Butyl 

Phosphate (TBP) when solvent is exposed to high levels of radiation. The 

degradation produces complex zirconium and ruthenium which subsequently 

contaminate both solvent and product streams. One of the purposes of the 

segregated solvent facilities is to remove the degradation end products and 

radioactive contaminants from the solvent.

Transfer Tanks

Seven transfer tanks provide intermediate pumping stations for solutions 

transferred from Building 221-H and are located in open basins, each containing 

a sump and a pump. The basins are considered a place of potential exposure to 

airborne radioactivity.

Sump Collection Tank

This tank receives radioactive waste condensate by steam jet from sumps and 

tanks related to the canyon air exhaust system and send it by jet stream to other 

tanks that feed the hot or warm canyon waste evaporators.

Recycle Sump

This sump collects the drainage and overflow for all 221-H tanks that contain 

contaminated or recycled liquids.

Recycle Vent System

The recycle vent system provides a means to filter contaminated air from tanks 

and vessels prior to discharge to the atmosphere.

Auxiliary Systems and Support Facilities

The process well water system is the only system of concern for these facilities. 

There are two cooling water streams that are of importance, the segregated 

cooling water stream and the recirculating cooling water stream. Both streams 

are continuously monitored for radioactivity and can be diverted into retention or 

seepage basins to prevent a release of radioactive material.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 40 of 188

Internal exposure - The radionuclide of concern for the A-Line is enriched uranium: 234U, 

235U, 236U, and 238U.

External exposure - The radionuclide of concern for the A-Line is natural uranium and 

the major contributor to the external dose will be photons and X rays from 234U, 235U, and 

236U.

The radionuclides of concern for all other facilities are 95Nb, 95Zr, 103Ru, 106Ru, and 137Cs.

A-19 Receiving Basin for Off-site Fuel (RBOF) and Resin Regeneration Facility (RRF)

The RBOF facility handles casks and fuels of various shapes, sizes, and content. Casks 

of spent fuel are received and washed in the cask basin to remove any dirt accumulated 

during transit. After the cask is loaded into the fuel basin and the loaded fuel basket 

within the cask is removed and transferred to a fuel transfer bucket, the fuel transfer 

bucket then moves the fuel underwater to either the inspection basin, disassembly basin, 

repackaging basin, or storage basin. There are facilities for inspecting the fuel by 

mechanical and optical means, disassembling fuel, cutting fuel by cross cutting or 

slitting, and weighing fuel before and after packaging or other alterations.

The RRF is capable of regenerating both anion and cation type resins. The mixed resin 

is first depleted with NaOH and then the cation and anion resins are separated after 

which they are regenerated, rinsed, and returned to the deionizer. Other operations 

performed at RRF include the removal and disposal of the filter cake from the portable 

100-Area filter, chemical cleaning of the filter stones in the portable 100-Area filter, and 

chemical cleaning of target slugs prior to processing in Building 232-H.

The irradiated fuel received at RBOF contains uranium and fission/activation products. 

Since the work is performed underwater, the inert gases that escape the fuel bundles 

are of primary radiological concern.

In addition to irradiated fuel, RBOF performs other operations such as basin purification, 

resin regeneration, and target cleaning. It was determined that tritium, fission/activation 

products, and plutonium could be present during these activities.

Internal exposure - The mixture of cation and anion resins in the RRF contains fission 

and activation products. The radionuclides of concern are 134Cs, 137Cs, and 60Co.

External exposure - The RRF receives a mixture of anion and cation exchange resins 

from the basin waste deionizer. The resins contain fission and activation products. 

A-20 Tritium Facilities

The Tritium Facility includes Buildings 232-H, 233-H, 234-H, and 238-H. Irradiated 

lithium targets are processed in the tritium facilities to recover, purify, and package 

tritium.

The radionuclides of concern for internal exposure are elemental tritium and tritiated 

water. In addition, the activation product, 65Zn is associated with the processing of 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 41 of 188

lithium targets in Building 232-H. The activation product 65Zn is assumed to be the 

primary radionuclide of concern for external exposure at the tritium facilities. 

A-21 M Area – Uranium Target Fabrication Facility, 313-M

The Uranium Target Fabrication Facility was designed and built to manufacture 

aluminum clad targets for irradiation in the SRS reactors.

Depleted uranium metal, received from the Fernald plant, was fabricated into target rods 

that were used for transmutation of 238U to plutonium.

There is no information on the distribution of radionuclides in the 313-M material, and the 

mixture given in Table A-21 is assumed. 

Internal exposure -The majority of the annual internal dose from this uranium mixture is 

delivered by 238U. 

External exposure - The majority of the external dose from the uranium mixture is 

delivered by 238U which decays to 234mPa which emits a high energy beta.

A-22 320-M, 322-M, and 341-M 

Process descriptions for these facilities are not available. However, target extrusion was 

conducted in Building 320-M, and Building 322-M was a Chemical and Metallurgical 

Laboratory. The M-Area facilities processed uranium, lithium, and aluminum into fuel and 

target components for the nuclear reactors. Processing included aluminum alloy 

formation, degreasing, etching, metal extrusion, hot-die size bonding, and nickel plating.

Uranium and/or enriched uranium are assumed to have been processed. The assumed 

distributions for depleted and enriched uranium are given in Table A-22. All 

radionuclides given in the tables are of concern for both internal and external exposures.

A-23 Fuel Fabrication facility, 321-M

The Fuel Fabrication Facility was designed and built to manufacture aluminum clad fuel 

elements for irradiation in the SRS reactors. In this process, enriched uranium metal is 

alloyed with aluminum, in concentrations required for reactor irradiation. The alloy is cast 

into hollow cylindrical ingots from which pre-extrusion billet cores are machined. These 

cores are extruded into logs, which are then machined into sections, encased in 

aluminum, and co-extruded into tubes. Cores for other types of tubes are fabricated and 

assembled into billets in other SRS facilities (Li-Al in Building 320-M and NpO2-Al in 

Building 235-F) before they are received at 321-M. The materials processed in Building 

321-M undergo the following processes: charge preparation; U-Al alloy storage; and UAl casting, core machining, billet assembly, and billet outgassing.

Three radionuclide sources were identified in Building 321-M: uranium, plutonium, and 

neptunium. The majority of the annual internal dose from the identified uranium mixture 

is delivered by 234U. Plutonium was also processed in 1980. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 42 of 188

The majority of the external dose from the uranium mixture is delivered by 234U, 235U, and 

236U.

When neptunium targets are prepared, photoemissions from neptunium and its 

daughters (i.e., 233Pa) present a much higher exposure potential than is normally 

encountered with uranium based targets.

A-24 S-Area – Defense Waste Processing Facility (DWPF) 

Radioactive wastes are stored in existing H-Area waste tank facilities in the form of 

settled sludge, supernatant liquid, and salt cake. The insoluble solids (sludge) is 

chemically treated and washed to remove aluminum in preparation for the feed to 

DWPF. The aqueous fraction undergoes treatment with sodium tetraphenylborate to 

precipitate the cesium and potassium salts. Also, a slurry of sodium titanate is added to 

absorb the soluble strontium and plutonium. The precipitate is washed and concentrated 

in preparation for feed to the DWPF. The aqueous product is subsequently blended with 

the formatted sludge feed and adjusted with glass formers. The adjusted feed is vitrified 

to a borosilicate waste glass in a slurry-fed joule-heated melter. The waste glass is 

encapsulated in stainless steel canisters and stored onsite in the glass waste storage 

building.

The radionuclide composition of the incoming DWPF feed stream may vary depending 

on which tanks of waste are being processed.

Two process streams represent the high and low ends of the range of 238Pu/90Sr 

concentrations.

Internal exposure - The sludge-slurry feed stream (Stream 1) has a specific activity of 56 

Ci/gal. The majority of the annual internal dose is delivered by 90Sr and 238Pu. 

The precipitate-slurry feed stream (Stream 2) has a specific activity of 37 Ci/gal. The 

majority of the annual internal dose comes from the 137Cs and 144Ce. 

External exposure - The majority of the external dose is delivered by 144Ce, 137Cs, 106Ru, 

90Sr, 238Pu, and 241Pu in the sludge-slurry feed (Stream 1) .

The precipitate-slurry feed stream has a specific activity of 37 Ci/gal with the majority of 

the external dose being delivered by the same radionuclides as from the sludge-slurry 

stream. 

A-25 Z Area

Waste tank supernate (low-level waste) is transferred from H Area to Z Area via pipeline. 

At Z Area, the salt solution (supernate) is mixed with cement, fly ash, and blast furnace 

slag; and the resulting grout is pumped to vaults where it sets into a saltstone monolith.

The radionuclide composition of the incoming feed stream may vary depending on which 

tanks are being processed.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 43 of 188

Two process streams have been identified on the basis of which radionuclides would 

deliver the majority of the dose and which radionuclides could be used for tracers. The 

process streams are the decontaminated salt solution and the saltstone.

Internal exposure - The majority of the annual internal dose from the decontaminated 

salt solution is delivered by 238Pu and 106Ru. The majority of the annual dose from the 

saltstone is delivered by 238Pu and 106Ru. 

External exposure - The majority of the external dose from the salt solution and the 

saltstone is delivered by 238Pu, 137Cs, 90Sr, and 106Ru.

A-26 A Area Building 773-A

The radionuclide distribution found in some of the various sections is not available, and 

the information provided is based on the type of work performed.

Building 773-A has many processes including the following:

Actinide Technology

This section provides technical support for the Pu and enriched uranium 

production facilities; Building 221-F, FB Line, Building 221-H, and HB Line. The 

material used here should be similar to that found in Building 221-F, FB Line, 

Building 221-H, and HB Line. The radionuclides of concern would include 

plutonium, neptunium, americium, and uranium. 

Analytical Development

This section provides analytical support for research programs and conducts 

research and development in the area of process control and analyzer 

development. The material used in this section may include fission products, 

uranium, neptunium, curium, and/or californium.

Defense Waste Processing Technology

This section provides support and conducts research and development in the 

area of waste processing. The material used in this section would be similar to 

the material processed in the S and Z area and should contain principally 

fission/activation products and plutonium.

Hydrogen Technology

This section is involved in development of hydride process technology and 

testing materials for use in corrosive and hydrogen environments. The type of 

radioactive material used in this section is unknown.

Interim Waste Technology

This section is developing the technology for the Defense Waste Processing 

facility (DWPF) and the Effluent Treatment facility (ETF). Work is performed on 

supernate processing, precipitation of strontium and cesium, and filtration and 

acid hydrolysis of the resulting sludges. The characterization of waste and the 

investigation of the treatment of effluents by ion exchange and reverse osmosis 

are also being developed. The material used in this section is assumed to be 

similar to the material processed in the S and Z Areas and should contain 

principally fission/activation products and plutonium.

Environmental Technology

This section processes environmental samples. The material used in this section 

may contain 129I and 14C.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 44 of 188

Laboratory Services

The areas of primary interest in this section include the High Level Cells and the 

F-Wing Experimental Area. The High Level Cells provide the shielding and 

confinement necessary for analysis and testing of highly radioactive material. 

The F-Wing Experimental Area has research and pilot scale production facilities 

for work with alpha and neutron emitters such as 239Pu, 244Cm, 243Am, and 252Cf.

Materials Technology

This section consists of a Robotics Development Laboratory, Powder Metallurgy 

Facility, On-site Uranium Reprocess facility, and the Hot Machine Shop. The 

purpose of the Powder Metallurgy Facility is to develop a new process for 

fabricating reactor fuel tube cores. The On-site Uranium Recycle Process facility 

is a prototype facility to assist in the design and operation of a plant scale Fuel 

Production Facility. The Hot Machine Shop provides a variety of functions and 

services including machining uranium metal. The material used in this section 

may include fission/activation products, tritium, and depleted and/or enriched 

uranium. 

A-27 Buildings 735-A and 735-11A

Buildings 735-A and 735-11A house the Radiological and Environmental Science 

facilities. It is assumed that mainly low-level environmental samples are present in these 

facilities.

A-28 Building 776-A Liquid Waste Handling Facility

Located in Building 776-A, this facility collects the aqueous waste from Building 773-A. 

Facilities in this building are provided to strain solids from the waste streams, collect 

waste in receiving tanks, sample and adjust the pH of the waste tank contents, and 

transfer the waste to tank trailers for disposal. The radionuclides are believed to be a 

composite of all the material that is used in Building 773-A operations.

A-29 D-Area Heavy Water Production and Reprocessing

A heavy water production plant in D Area began operation early (1952) in SRS history to 

concentrate heavy water from Savannah River water to moderate and cool the Site’s 

reactors. The facility stopped production in 1981 because there was a sufficient supply 

of heavy water.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 45 of 188

2.0 OCCUPATIONAL MEDICAL DOSE

SRS required pre-employment and annual physical examinations as part of their 

occupational health and safety program. These medical examinations typically included 

diagnostic chest x-rays. The doses from these diagnostic x-ray procedures depended 

not only on the characteristics of the x-ray machine and the procedure used, but also on 

the frequency of the examination. 

2.1 Examination Frequencies

The frequency of exams differed significantly over the years. Occupational x-rays may 

have occurred more frequently than on the schedule indicated in Table 2.01. This 

information will be recorded in the file provided by DOE. All occupational x-rays in the 

worker file are to be included in the reconstruction of occupational dose.

Table 2.01 lists the frequencies of chest x-rays for different age groups through the 

years and also identifies specific groups of workers. Specific organ doses to be 

attributed for PA chest X rays calculated on the basis of the dose conversion factors 

found in ICRP Publication 34 are given in Attachment B, Table B.01. Organ doses from 

lateral chest radiography have been estimated at 2.5 times greater than those from the 

corresponding PA doses based primarily on the greater mAs exposure per radiograph 

and the somewhat smaller Source to Skin Distance (SSD). For organs not listed in ICRP 

Publication 34 but specified in the IREP code, doses were determined by analogy with 

anatomical location (Table 2.04). Thus IREP code organs in the thoracic cavity, but not 

mentioned in the ICRP Publication 34 were assigned the same dose as the lungs; doses 

to the organs in the head and neck were assigned the same dose as the thyroid. The 

head and neck organ dose estimates (i.e. eye/brain), should be somewhat greater than 

doses actually incurred (hence claimant favorable), because of geometry considerations 

and at least in the case of brain, because of attenuation by the bony cranium. To ensure 

claimant favorability in the view of the variations in organ dose described in ICRP 

Publication 34 (p. 51), the doses for females, which are slightly higher than those for 

males, were used. 

2.2 Equipment and Techniques

Although the Savannah River Site (SRS) medical practices are assumed to have 

followed the adoption of standards of radiology practice during the 1930s and 1940s to 

minimize dose to the patient, there is the potential for significant dose from occupational 

medical x-ray examinations depending upon the type of equipment, the technique 

factors, the number of photo-fluoroscopic examinations typical in the early years 

(Cardarelli 2002) and the number of radiographic examinations. SRS medical records 

include notations in individual worker files regarding both the date and the purpose of the 

of x-ray examinations. 

X-ray organ dose estimates for occupational x-rays administered at the SRS are made 

for Type I equipment (used from 1950 through 1970), Type II equipment (used from 

1971 to July 1985), Type III equipment (used from August 1985 to April 1999), and Type 

IV equipment (used from June 1999 through 2002). A description of the x-ray equipment 

used at the SRS is included in Table 2.02. The specific technique factors for these 

machines are in Table 2.03. Since no technique factors were identified by the SRS for 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 46 of 188

Type I equipment, organ dose estimates for Type I were made by multiplying the organ 

dose estimates from Type II equipment by 2.5. This approach is reasonable when 

compared to other DOE sites (e. g. Hanford) where more information is available for xray equipment from the early time period. 

For both posterior/anterior and lateral views, a standard source to image distance (SID) 

of 72 inches (183 cm) was maintained for all types. Additional information indicated that 

all of the x-ray machines were single phase and that there was no air gap between the 

patient and the film. It is assumed that the Type II-IV equipment had at least 2.5 mm Al 

total filtration (see Table 3.1 of NCRP 102) and therefore a HVL of 3.3 mm Al eq. (see 

Table B.2 of NCRP 102). Tables B-01 and B-02 list organ doses for PA 14” x 17” chest 

films.

Table 2.01 Frequency of Occupational Chest X-rays at the SRS

Time Period Frequency Comment

1950 – 1988 Annually All employees.

Annually (1) Construction asbestos workers.

1989 – 1993

Annually Employees over 50 years old.

Biennially Employees 40 to 49 years old.

Every 3 years Employees under the age of 39.

Biennially Asbestos workers.

Annually DOE, USFS and WSI employees

1994 – 1998

Biennially Employees over 50 years old.

Every 3rd year Employees 40 to 49 years old.

Every 5th year employees under the age of 39

Biennially Asbestos workers.

Annually DOE, USFS and WSI employees

1995 Biennially Pensioners with documented uptake of 5 rem

1999 – 2002

Every 5th year Employees in a health surveillance program

Annually DOE, USFS and WSI employees

Biennially Asbestos workers.

(1) The annual chest x-ray was a minimum received by construction and asbestos workers. 

Individual workers may have received chest x-rays as often as every 6 weeks. This would be 

recorded in the file provided by DOE on the individual worker.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 47 of 188

Table 2.02 Description of the X-ray Equipment used at SRS

Technique Time Period Equipment

Type I 1950-1970

Dupont 2DC Safety Screens, Dupont Cronex 7 film, 

Picker x-ray Tube, No grid, Manual/Hand processing, 

Manual collimator.

Type II 1971-7/1985

Dupont Daylight Hi-Speed Screens, Dupont Cronex 

7Film, Picker x-ray Tube, Stationary 12:1 Grid, PAKO 

3-Minute Film Processor, Manual Collimator.

Type III 8/1985-5/1999

Dupont Daylight Cronex 10 TL Film, Quanta III 

Screens, TecRad Manual Collimator, Technomed 

Recipromatic Upright Bucky 12:1 Grid, Kodak M6AW 

90 sec. x-ray Film Processor, Eureka x-ray Tube.

Type IV 6/1999-2002

AGFA Cronex 10TK Film, AGFA Curix CassettesOrtho Regular Screens, AGFA Multiloader Processor, 

TecRad Manual Collimator, Eureka x-ray Tube Rad 68, 

Technomed Recipromatic Upright Bucky 12:1 Grid.

Table 2.03 Technique factors Used for Each Type of X-ray Equipment

Machine View Current

(mA)

Voltage

(kVp)

Exposure Time

(sec.)

Type I

Techniques unknown. No records were kept with regard to the technique 

factors used during this time period. The Radiologic Technologists employed 

during this period are all deceased; therefore, no known resources exist.

Type II PA 300 110-120 1/30

Type III PA 300 120 1/40 

Type IV PA 300 120 1/40 

Note: PA indicates a posterior/anterior view, the average PA chest measures 26 cm.

the average Lat. chest measures 34 cm.

Although the SRS has not provided any information on early medical X rays, there is 

indication that a mobile X-ray unit was used that may have resulted in higher doses. 

DPSP-57-01-11 indicates that an evaluation of the mobile X-ray unit was conducted and 

that full size chest radiographs were recommended to reduce personnel exposures. 

This implies that higher exposures may have been delivered by this unit, which may 

have been a photofluorographic machine.

 

 

2.3 Organ Dose Calculations

ICRP Publication 34 (1982) provides tables of average absorbed dose (mGy) in selected 

organs for selected x-ray projections at 1 Gy entrance kerma (i.e., air kerma without 

backscatter), for selected views (including PA), and for selected beam qualities (i.e., 

various HVLs). These tables provide the basic dose conversion factors for converting air 

kerma to organ dose. Air kerma was obtained from Table B.3 of NCRP Report 102 

(1989). The average air kerma rates for the different machines are calculated using the 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 48 of 188

cGy per mAs provided in NCRP No. 102 (1989) for specific voltage, current, phase of 

the machine, and distance to the film. 

Finally, the SRS organ doses are found by multiplying the ICRP 34 organ dose 

conversion factors by the modified air kerma factors. The resulting SRS X ray organ 

doses for all machines are in Tables B.01 and B.02. The doses are shown in units of 

rem, assuming a quality factor of 1.0 for X rays. SRS records will indicate the view and in 

most cases only one view was taken per medical examination. 

The following formulas were used to calculate the doses in Table B.01.

Based on the techniques in Table 2.03, the mAs for the types of equipment were 

calculated for each view:

Current (mA) x Exposure Time (sec) = Current for View (mAs) (2-1)

Example for Type II PA view: 300 mA x 1/30 s = 10 mAs 

The air kerma rate for 120 kVp was determined to be 0.3 cGy per 100 mAs (see Table 

B.3 of NCRP 102) and the air kerma was calculated after converting the rate to air 

kerma per mA.

Current for View (mAs) x Corrected Air Kerma Rate (cGy/mAs) = Air Kerma (cGy) (2-2)

Example for Type II PA view: 10 mAs x 0.003 cGy/mAs = 0.03 cGy

The air kerma was corrected for the thickness of the chest (26 cm) and for distance 

between the chest and the plane of the film (5 cm) to obtain the air kerma at skin 

entrance.

Air kerma at 183 cm x SID squared ÷ SSD squared = air kerma at skin entrance (2-3)

Example for Type II PA view: 0.03 cGy x (183 cm)2

 ÷ (152 cm)2 = 0.043 cGy

Air kerma at skin entrance was multiplied by the dose conversion factors in Table A.2 

through A.8 of ICRP Publication 34 for PA chest and HVL of 3.5 mm Al eq.

Air Kerma (cGy ) x Dose conversion factor = Dose for View (cGy) (2-4)

Example for Type II PA view, dose to thyroid: 

0.043 cGy x 62 mGy/Gy x 1 Gy/100cGy x 1 rad/10 mGy = 2.7 E-3 rad

For Type 1 equipment used prior to 1970, no collimation was assumed. Therefore, 

organs not normally in the primary beam for a PA chest were included in the primary 

beam by using ICRP 34 organ dose conversion factors for procedures where those 

organs would normally be included in the primary beam. For example, assuming no 

collimation, the ovaries would be in the primary beam, and dose conversion factors for 

the abdomen were used. Abdomen dose conversion factors were used for ovaries, 

testes, embryo, and their analogues, cervical spine dose conversion factor was used for 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 49 of 188

the thyroid and its analogues, and the PA skull dose conversion factor was used for the 

eye/brain.

Table 2.04 Analogues for IREP Organs Not Included in ICRP 34

Anatomical 

Location

ICRP 34 Reference 

Organ 

IREP Organ

Analogues

Thorax Lung Thymus

Esophagus

Stomach

Bone Surface

Remainder organs

Abdomen Ovaries Liver/Gall Bladder

Urinary/Bladder

Colon/Rectum

Uterus

Head and Neck Thyroid Eye/Brain

2.4 Uncertainty

Error, defined as deviation from the correct, true or conventionally accepted value of a 

quantity, and uncertainty, defined in terms of the potential range of a stated, measured, 

assumed or otherwise determined value of a quantity, provides an indication of the 

confidence of the dose estimates. Error implies knowledge of what the correct or actual 

value is, which is, of course, not known. Hence a more appropriate term is uncertainty, 

which is expressed in terms of a confidence level, e.g. 99% (i.e. that the correct or true 

value, although not actually known, has a 99% probability of falling within the range 

cited) and includes both precision or reproducibility of the measurement and accuracy, 

or how close the measurement or estimate of dose comes to the actual or correct value.

Although in theory a large number of factors can introduce uncertainties or affect the xray machine output intensity and dose to the patient, in practice only four factors can be 

reasonably considered to have an impact on dose uncertainty. These are 1) variation in 

applied kilovoltage, 2) variation in beam current,3) variation in exposure time, and 4) 

distance from the patient to the source of the X rays (SSD). The influence of such other 

factors as use of screens, grids, reciprocity failure, film speed and development, while 

potentially variable, would not affect the beam output intensity.

For a given set of machine settings and parameters, x-ray output should theoretically be 

constant and unvarying. However, this is not true in practice; although output is 

essentially constant unless focal spot loading occurs such as might be the case when 

the power rating of the machine is exceeded. It is unlikely that power ratings were ever 

exceeded since so doing would be difficult to achieve in practice and would result in 

damage to the x-ray tube. However, even with the use of so-called constant voltage 

transformers to control line voltages, slight variations may occur in line voltage input or 

other internal voltages, which in turn could alter the kVp of the output beam. In general, 

for a given kVp setting, variation in kVp falls within ± 5% of the machine setting. Since 

as noted above beam intensity is approximately proportional to the 1.7 power of the 

kilovoltage, this translates to an uncertainty of approximately ± 8.7%with respect to 

output beam intensity in the 100 to 120 kVp range used for diagnostic radiographs at the 

Savannah River Site . For conservatism, this is rounded up to ± 9%.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 50 of 188

Similarly slight variations in tube current are normal; as a tube ages, or heats up from 

usage, tube current may change and typically will drop. Hence, all other factors 

remaining constant, beam intensity will be reduced, and in direct proportion to the 

change in tube current. Typically, the reduction in beam output from current variation is 

not more than a few per cent under normal operating conditions; large decreases in 

beam output will be readily detected and result in maintenance on the machine to 

restore the output, or, as a temporary stopgap measure, increase in the current or kVp to 

provide the necessary intensity for proper radiography. There is no evidence to suggest 

that these stopgap measures were ever necessary or applied at the Savannah River 

Site. For a given kVp setting, output of the beam is a function of the tube current, which 

in turn is measured by a milli-ammeter on the machine and measures average tube

current. The measurement is subject to uncertainties, and in addition there may be 

minor changes in output as the tube heats up from normal usage. These variations are 

typically small, and hence uncertainty in beam output attributable to current variation has 

been estimated at ± 5%.

Another parameter that has potential to affect the dose, perhaps significantly, from a 

diagnostic radiograph relates to the time of exposure. This can be readily understood by 

noting that a full wave rectified machine produces 120 pulses per second of x-rays. For 

an exposure time of 1/20 of a second, only six pulses would result. A small error in the 

timer that resulted in a change of only ± 1 pulse would correspondingly affect the output 

by ± 17%; for an exposure time of 1/30 of a second, the change in output corresponding 

to a deviation of ±1 pulse is ± 25%. Early mechanical timers were notoriously 

inaccurate, although timer accuracy improved significantly with the introduction of 

electronic timers. However, once again for conservatism, uncertainty in beam output 

attributable to timers will be assumed to have an upper limit of + 25%. 

The final factor that is likely to affect patient dose relates to distance from the source of 

the x-rays, which is a determinant of the entrance skin exposure. For a given individual, 

the SSD will be determined largely by the thickness of the patient, and how accurate the 

positioning is. For a typical patient, this variation in SSD is estimated at no more than a 

few centimeters, with an upper limit of perhaps 7.5 cm. Using inverse square, this 

indicates an uncertainty of ± 10 % from this source.

There are two approaches to determination of the combined uncertainty from the above 

four potential sources of uncertainty. The first, and most conservative in that it gives the 

greatest range, would be to assume that the uncertainties are additive, which would give 

an uncertainty range of up to 9 +5+25+10 = 49. However, a more reasonable approach 

would be to assume that the uncertainties are in fact random, and to compute the 

statistical root mean square (RMS) value. The RMS value is simply the square root of 

the sum of the squares, and computes as ± 28.7%. Thus, for any individual entrance 

skin exposure (ESE) or derived organ dose, an uncertainty of ± 30 % at the 99 % 

confidence level may be assumed; for further conservatism it may be appropriate to 

assume that errors are all positive, and only the + 30% should be used. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 51 of 188

3.0 OCCUPATIONAL ENVIRONMENTAL DOSE

The occupational environmental dose refers to the dose received by workers outside of the 

facilities. These doses can be internal and external depending on the characteristics of the 

individual radionuclides. While most radionuclides when inhaled would represent a dose to 

various organs in the body, the noble gases would only represent an external dose. These 

radionuclides are addressed in the following sections.

3.1 Internal Dose from On-Site Atmospheric Radionuclide Concentrations

The internal dose for workers outside of the facilities is determined from the air 

concentrations resulting from the individual facility releases, from ground level releases 

and from the resuspension of radioactive materials in the soil.

Unmonitored workers can be exposed to occupational doses internally from on-site 

releases into the air, from the resuspension of radioactive materials in soil, and 

externally from ambient radiation and releases of radioactive noble gases into air.

3.1.1 On-Site Releases to Air

Although the annual and semiannual environmental reports contain air monitoring data, 

these data were determined to be of limited value because of the confounding effects of 

nuclear weapons testing, lack of spatial and radionuclide resolution, and insensitivity of 

sampling methods to radionuclides important to dose. Therefore, methods were 

developed to use well- documented source terms developed by others (Cummins 1991; 

CDC 2001), coupled with historical meteorology, to estimate radionuclide-specific 

airborne concentrations for 3

H, 131I, 41Ar, 238Pu/239Pu/240Pu, and 234U/235U/238U. These 

radionuclides were determined to account for more than 95 percent of the potential 

missed dose from inhalation and submersion pathways. Air concentration calculations 

included spatial and release height considerations for source terms from two processing 

areas (F- and H-Areas), the heavy water extraction area (D-Area), and five reactor areas 

(C-, K-, L-, P-, and R-Areas). The 50th and 95th percentile air concentrations were 

estimated for the eight source term areas, seven other populated areas (shown in Figure 

3.1.1-1), and a site average. These calculated air concentrations were the used to 

derive annual intakes based on an assumed individual ventilation rate of 2,400 cubic 

meters per year. In addition, the geometric standard deviations (GSD) of these intakes 

were estimated based on the 95th percentile source terms (Rollins 2003) using the 

following relationship.

)()( 65585.1

1

50 th percentile

95 th percentile GSD = (3-1)

These annual intakes and GSDs, by year of operation, are presented in Attachment C in 

Tables C-01 through C-17. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 52 of 188

Figure 3.1.1-1 The SRS showing the reactor (C-, K-, L-, P-. and R-Reactors), and processing 

areas (F Area and H Area) in a rough circle toward the center of the Site, and D Area and A 

and M Areas near the Site perimeter. Also shown are DWPF (S Area), Saltstone (Z Area), 

Central Shops, Heavy Water Extraction (D Area), U.S. Forest Service Headquarters, and the 

Railroad Classification Yard.

Atmospheric dispersion was modeled by developing atmospheric dispersion (X/Q) 

values using 5-year average joint frequency distributions presented in the SRS 

environmental reports between 1975 and 1999 (see reference section). These data 

were converted to X/Q tables for elevated and ground-level releases, using the 

Multimedia Environmental Pollutant Assessment System (MEPAS) 4.0 computer code 

(PNNL 1996) and plotted using an Excel® spreadsheet. To assure that atmospheric 

R Reactor

P Reactor

L Reactor

K Reactor 

C Reactor 

 F Area

H Area

D Area

M, A Areas

Central 

E Area

S Area

Z Area

RR Yard

Forest Service HQ
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 53 of 188

dispersion would not be overestimated and resultant air concentrations underestimated, 

the highest X/Q values in any direction over the 25-year 

period were chosen as representative for all years of operation. Studies performed at 

the SRS have indicated that the use of the highest X/Q values would not significantly 

over or under estimate actual atmospheric dispersion for all years of operation (Harvey 

2001, Hamby 1991). Expressions were developed for ground level and elevated 

releases and used to estimate atmospheric dispersion based on distance from the 

source. These expressions were developed using the Excel® spreadsheet for all sourceterm locations (reactor and separations areas) from 500 m out to a distance of 80 km.

For ground level releases (5 m):

Y = 1.0146 X-1.8809 (3-2)

For elevated releases (60 m):

Y = 0.0013X-1.1739 (3-3)

Where

Y = Atmospheric dispersion coefficient, s per m3

X = Distance from source in meters.

Distances between the populated areas were estimated using geographical information 

provided by Garmin (2001) and are presented in Table 3.1.1-1. These distances were 

used in the X/Q expressions and coupled with the adjusted source terms to estimate 

cumulative air concentrations for all 15 population areas evaluated (Rollins 2003 - TBD). 

Table 3.1.1-1 Distance, in Meters, from Source Areas to Population Centers at SRS.

Source

Location

A

Area

C

Area

D

Area

F

Area

E

Area

H

Area

S

Area

Z

Area

R

Area

RR

Yard

P

Area

L

Area

K

Area

Forest

Service CS

C Area 11500 500 7500 3800 4100 5400 6600 7500 9400 12500 9400 6600 4300 13500 2200

F-Area 8100 4000 10600 500 1600 3500 3900 4700 9100 13900 11300 9700 8300 9700 6500

H Area 10500 5500 12900 3500 1900 500 1200 2000 5700 11100 8900 8800 8900 8800 6400

K Area 15800 4300 6300 8200 8200 8800 10100 10800 10500 11300 8200 3900 500 17500 6900

L Area 17800 6600 10150 9600 8900 8600 10100 10500 8100 7400 4400 500 3900 17600 7600

P Area 19100 9400 14400 11100 9900 8800 9900 10100 5300 3100 500 4400 8100 16900 9600

R Area 15900 9400 16300 9100 7500 5700 6200 5900 500 6100 5300 8300 10500 12000 9400

For radionuclides potentially important for internal dose reconstruction, the average and 

maximum annual intakes and respective GSDs for all populated areas are provided in 

Tables C-16 and C-17, respectively. Area-specific annual intakes with respective GSDs 

are provided in Attachment C (Tables C-01 through C-15) for use where more detailed 

information is required. The values in these tables were derived from calculated 

concentrations for the major populated site locations based on historical on-site 

meteorology, distance from release points, and emissions data. 

The values in Tables C-01 through C-15 are estimates of annual intakes derived from 

the calculated outdoor air concentrations at SRS locations and, as such, may not be 

appropriate for calculating organ doses for workers located inside buildings. However, 

these values likely represent the upper bound of potential intakes for an individual 

working in outside areas where individual monitoring was not required. It is assumed 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 54 of 188

that these intakes occurred uniformly in the year listed, so for partial years of 

employment, the intake may be proportionally adjusted by the fraction of the year 

worked. In addition, these intakes were derived based on an assumed ventilation rate of 

2,400 cubic meters per year. Therefore, adjustments may be required based on the job 

classification depending on the individual performing light work or heavy work for which 

the respiration rate would be 1.3 m3

/h or 1.7 m3/h respectively.

For purposes of dose estimation, radioactive decay was not considered. This is 

considered a claimant-favorable assumption because it would tend to overestimate the 

actual air concentrations and, thus, the resultant radiation dose. (see Section 3.3 for 

other uncertainty discussions

In performing dose calculations for unmonitored individuals, three situations may arise;

1. If a worker spent significant time in areas for which radionuclide intakes were 

calculated and the worker was unmonitored the data in Tables C-01 through C-15 may 

be used to estimate intakes. These tables show the maximum annual intakes at the 50th

percentile and respective GSDs for 3

H, 131I, Pu, and U, in each SRS area by year.

2. If the individual worked in multiple areas on the site, the site average annual intakes in 

Table C-16 should be used. Table C-16 lists the site-wide average annual intakes, which 

is recommended for performing dose reconstructions to reflect the most likely annual 

intake to an unmonitored individual. For uncertainty estimation, the GSD (based on the 

95th percentile radionuclide source term may be used to define the lognormal 

distribution.

3. Table C-17 shows the maximum annual intakes of these radionuclides at any SRS 

area by year. To perform internal dose reconstructions for unmonitored individuals when 

a specific work location cannot be identified for an individual, maximum annual intakes in 

Table C-17 (derived from Tables C-01 through C-15 in Attachment C) for each year of 

employment can be used to estimate the dose to the organ of interest for radionuclides 

that are most important to dose (Rollins 2003). The use of the maximum values would 

represent a claimant-favorable dose. The decision to use these maximum annual intakes 

to estimate unmonitored organ doses should be made cautiously because the worker’s 

intake could be significantly overestimated

These intakes are to be used by internal dose reconstructors along with appropriate 

values for breathing rate to derive the annual intakes for the significant dose-contributing 

radionuclides at SRS. Annual intakes provided in Table C-01 through C-17 were derived 

from the calculated air concentrations assuming a ventilation rate of 2,400 cubic meters 

per year. These intakes apply uniformly to the year listed, so for partial years of 

employment, the intake may be proportionally adjusted by the fraction of the year 

worked.

The following sections discuss details of specific radionuclides of importance for 

calculations of airborne concentrations
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 55 of 188

Tritium

Tritium is one of the principal nuclear materials produced at the SRS. It is the heaviest 

and only radioactive isotope of hydrogen with a physical half-life of 12.5 years. Tritium 

processing operations have been performed at the SRS facility since October 1955, 

beginning in Building 232-F. In July 1957, the process was moved to Building 232-H and 

operations doubled by 1958. 

Tritium is produced naturally by cosmic ray interactions and as a fission product in 

nuclear reactors. The nuclear reactions that produce tritium are:

• Fission of lithium by neutron irradiation of targets

• Reaction of neutrons with the heavy water moderator used in reactors

• Ternary fission of transuranic elements in the reactor fuel and targets.

At SRS, tritium was originally produced as a by-product of producing plutonium in the 

reactors. Lithium-aluminum control rods were used in R-Reactor and other reactors from 

the first day of operation to produce tritium and control reactor power. When additional 

tritium was needed beginning in 1955, enriched uranium was used in the reactors, which 

allowed for increased loading of lithium-aluminum targets for tritium production.

Gaseous tritium at room temperature tends to form elemental tritium (HT) by reaction 

with gaseous hydrogen. Tritium closely follows the reactions of ordinary hydrogen when 

it exists as tritiated water or vapor (T2O); tritium oxide [HTO]) is taken into the body quite 

easily, resulting in a higher factor of risk. Tritium is released from SRS in both elemental 

and oxide forms.

The key processes that have led to tritium releases at the SRS are:

• Reactor operations (100 Areas)

• Recovery of transuranic elements in the separations facilities (200 Areas)

• Recovery of tritium in the tritium processing facilities (200 Areas)

• Laboratory research area (700 Area)

• Heavy water rework facility. (400 Area)

The majority of the tritium releases came from the reactor facilities (100 Areas) and the 

tritium processing facilities (200 Areas). For purposes of on-site dose reconstruction, 

only releases from these areas are considered because these releases account for the 

vast majority of the total tritium released to the atmosphere. Additional information, 

including histories and descriptions of operations at these facilities and the reconstructed 

atmospheric tritium source terms from 1955 through 1992, can be found in Savannah 

River Site Dose Reconstruction Project Phase II: Source Term Calculation and Ingestion 

Pathway Data Retrieval, Evaluation of Materials Released from the Savannah River Site, 

RAC Report No. 1-CDC-SRS-1999 (CDC 2001, Chapter 4.1)(hereinafter referred to as 

the RAC Report). Source terms from 1993 through 2001 were obtained from the SRS 

annual environmental reports (see reference section for citations). The 50th percentile 

values were assumed to be those presented in the annual environmental reports. The 

95th percentile values were estimated by dividing the 50th percentile yearly value by the 

smallest value of the ratio of the 50th to the 95th percentile values for all of the years 

reported in the RAC Report (CDC 2001). This method provides reasonable assurance 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 56 of 188

that the 95th percentile values for 1993 through 2001 are not underestimated (see Rollins 

(2003) for details).

The uncertainty analysis applied to reported releases is described in Attachment E of the 

RAC Report (CDC 2001). The adjusted tritium source terms and uncertainty analyses 

(CDC 2001, Chapter 4.1) used to calculate atmospheric concentrations attempted to 

account for the insensitivity of the historical sampling methods and their inability to 

remove tritium oxide and elemental tritium effectively from the atmosphere. Table 3.1.1-

2 presents a description of the monitoring equipment used to monitor atmospheric tritium 

releases.

Table 3.1.1-2 Overview of Tritium Monitors at SRS (Table E-1, CDC 2001, p. E-2)

Monitor Facility Time period in use

Dehumidifier Reactors 1954–shutdown

Silica gel Reactors and 

disassembly areas

during 1954–1958 or until shutdown in 

some cases

Kanne ionization 

chambers Tritium facilities 1954–present

Stack tritium monitor 

(STM) Reactors 1970–1988

Berthold tritium monitor 

(BTM) Reactors 1988–present

Stack monitor integrator 

(SMI) Tritium facilities 1974–present

FORMS (STM with 

dehumidifier)

Tritium facilities and 

reactors 1985–present

In the reactor areas, the primary atmospheric release points are the 200-foot stacks and 

ground-level evaporation from the disassembly basin water. An additional source of 

atmospheric tritium release is the ground-level evaporation of water purged from the 

disassembly basin to a seepage basin. Tritium in the form of DTO (from the moderator) 

is released to the ventilation system and, therefore, to the stack by evaporation of 

process water exposed to air flowing through the process area. During reactor 

operations, D2O containing DTO is evaporated from small leaks in pipe flanges, valves, 

and from the D2O process water exposed in the control rod guide tubes. To account for 

the ground-level releases, the conservative assumption was made (based on The RAC 

Report (CDC 2001, p. 4.1-8)) that 8% of the total tritium released from the reactor areas 

was from ground level. This additional amount was added to the adjusted source terms 

reported in the RAC Report (CDC 2001) and modeled as ground-level release. All 

releases from the processing facilities were modeled as elevated (200-foot) releases.

Although between 30% and 50% of all tritium released from SRS is thought to be of 

elemental form (RAC Report (CDC 2001, pp. 4.1-15, 16)), the annual intakes presented 

in Tables C-01 through C-15 did not attempt to account for this. Therefore, the dose 

reconstructors should assume that the annual intakes consist entirely of the more 

biologically effective form of tritium oxide.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 57 of 188

Radioiodine

This section discusses the releases of the most important radioiodine isotopes (131I and 

129I) to the atmosphere from facilities at the SRS. The historic measurements at the 

facilities were interpreted in the context of current knowledge of radioiodine chemistry 

and the effectiveness of various sampling techniques (RAC Report (CDC 2001, Chapter 

4.2-2)). These analyses that considered the chemical form, sample line losses, and 

efficiency of sampling methods resulted in revised estimates of radioiodine releases. 

These revised estimates, or radioiodine source-terms, were used to calculate air 

concentrations and annual intakes at 15 populated areas within the SRS. A detailed 

discussion of methods used to develop the revised source terms and the uncertainty 

analyses used to develop the 95th percentile values is provided in the RAC Report (CDC 

2001, Chapter 4.2). These 95th percentile values were used to derive the geometric 

standard deviations shown in Tables C-01 through C-17. An overview of the radioiodine 

sampling and monitoring methods used at the SRS is provided in Table 3.1.1-3.

Table 3.1.1-3 Overview of Radioiodine Sampling and Monitoring Methods (RAC Report 

(CDC 2001, Table 4.2-2))

Period Method of

sample collection Comments on reported results

Dec 1954 to 

Oct 1956 NaOH scrubber

Only reactive forms were measured; system was not 

considered reliable; no corrections for line loss were 

made

Oct 1956 to 

Sept 1961

AgNO3-impregnated 

filters

Only reactive forms were measured; unknown 

variability in collection efficiency; no corrections for 

line loss were made

Sept 1961 to 

Sept 1965 Charcoal cartridges

All chemical forms were measured; attention paid to 

collection efficiency variations; no corrections for line 

loss were made

After

Sept 1965 Charcoal cartridges

New monitoring system with higher flow rate; all 

chemical forms were measured; attention paid to 

performance; no corrections for line loss were made

Revised radionuclide source terms were developed for both organic and inorganic 

chemical forms. Although organic radioiodines comprised a large fraction of the total 

radioiodine released to the atmosphere, for purposes of on-site dose reconstruction, the 

calculated annual intakes presented in Tables C-01 through C-17 are based on total 

radioiodine released without distinction to chemical form. This assumption is deemed 

appropriate because the major differences in the biological effectiveness of the chemical 

form occur primarily in the air-grass-cow-milk pathway, which is not considered for onsite dose reconstruction.

The fuel processing facilities in F Area and H Area were the sources of the highest 

releases of radioiodines, which were discharged through tall (60-m) stacks. However, 

for completeness, elevated releases of radioiodines from the reactor areas were also 

included in the air concentration calculations and resultant annual intakes. The largest 

reported releases occurred in 1956. Releases during the years 1955–1961 were 

significantly higher than the releases that occurred between 1962 and 1971. Those 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 58 of 188

latter releases were, in turn, markedly greater than releases in the later 1970s and 

1980s.

The revised source terms from the processing facilities were provided for the combined 

releases from F and H Areas (RAC Report (CDC 2001, Chapter 4.2)). To improve 

spatial resolution of the calculated annual intakes, these combined source terms were 

partitioned into individual area releases by comparing each area’s uranium through-put 

to the total uranium through-put (RAC Report (CDC 2001, Chapter 4.2 data files)) for 

both reprocessing areas. 

Similarly, revised source terms were provided for combined releases from the five 

reactor areas. Since there were no quantitative measurements of iodine releases to the 

atmosphere from the reactor areas prior to 1972, revised estimates were developed 

based on radioiodine measurements in reactor liquid effluents and later measurement of 

ratios of atmospheric to liquid releases. To improve spatial resolution of the on-site 

calculated air concentrations and resultant annual intakes, individual reactor 

atmospheric releases were estimated based on ratios of reactor-specific radioiodine 

liquid releases to the combined source terms for all reactors.

Annual intakes of 129I were not calculated because the quantity of 129I released to the 

atmosphere was several orders of magnitude lower than that of 131I and therefore not 

deemed important to on-site dose reconstruction. 

Plutonium and Uranium

Plutonium and uranium releases to the atmosphere represent the greatest potential 

radiological risk of all the alpha-emitting radionuclides in the SRS environment. 

Although releases of these radionuclides were reported for the reactor areas, M Area 

facilities, and the Administration Area (A Area), the majority of the releases were 

reported for the two chemical separations areas canyon facilities (221-F and 221-H). 

Because releases of plutonium and uranium from F- and H- Areas represent more than 

97% of the total atmospheric releases of all alpha-emitting radionuclides, air 

concentrations and resultant annual intakes were calculated only for uranium and 

plutonium released from the chemical separations areas. The slight underestimation of 

air concentrations and potential doses resulting from this simplifying assumption is 

considered to be well within other identified uncertainties.

In general, from 1955 through 1989, the source terms reported by Cummins (1991) for 

plutonium and uranium were determined to be the best available data (RAC Report 2001 

(CDC 2001)). Uncertainties were estimated (RAC Report (CDC 2001, Chapter 4.4)) 

based on sample collection and counting procedures, and estimated possible sample 

losses during transmission through sampling probes and lines. These reported values 

and their estimated uncertainties were used to calculate the 50th percentile annual 

intakes and geometric standard deviations presented in Tables C-01 through C-17. 

Source terms from 1990 through 2001 were obtained from the SRS annual 

environmental reports (see reference section for citations). The 50th percentile values 

were assumed to be those presented in the annual environmental reports. The 95th

percentile values were estimated by dividing the 50th percentile yearly value by the 

smallest value of the ratio of the 50th to the 95th percentile values for all of the years 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 59 of 188

reported in the RAC Report (CDC 2001). This method provides reasonable assurance 

that the 95th percentile values for 1989 through 2001 are not underestimated.

To simplify the analysis and because isotopic data are limited, calculated annual intakes 

were developed for total releases of uranium and plutonium. Isotopes assumed to be 

present in the mix include 238Pu/239Pu/240Pu and 234U/235U/238U. For purposes of dose 

reconstruction, the isotope of plutonium and uranium contained in the mix that results in 

the maximum organ dose should be selected and assumed to be present at 100%. This 

assumption will result in a small, but claimant-favorable, overestimation of the actual 

dose.

3.1.2 Annual intakes from Resuspension

Soil sampling and analysis were not routinely performed at the Savannah River Site 

during the period of greatest atmospheric releases (from 1955 through the late 1960s). 

However, several special surveys were performed between 1958 and 1970, but methods 

for soil sampling and analysis were not standardized throughout the DOE weapons 

complex until the early 1970s (RAC Report (CDC 2001, Chapter 12.2, p. 12.2-1)). As of 

1973, a laboratory had been dedicated to soils analysis and the necessary field and 

laboratory equipment had been purchased or fabricated. 

Soil analyses conducted in the 1970s and 1980s indicated no measurable SRS-released 

plutonium at off-site locations. However, elevated levels of certain radionuclides, 

including plutonium, were detected at several locations near the release points in both Fand H-Areas. Careful review of these data over several years concluded that the 

measured soil concentrations in F and H Areas were due almost entirely to deposition 

during the years of the greatest releases. Since the great majority of 239

Pu was released 

from F Area in a single year, 1955, and the majority of 238

Pu was released from H Area 

in 1969, the measured soil concentrations for 239

Pu could be conservatively assumed to 

have persisted in F Area since 1955 and for 238Pu in H Area since 1969 (RAC Report 

(CDC 2001, p. 12.2-7)). Because the atmospheric releases of nonvolatile radionuclides 

from other on-site locations were much less than those from F and H Areas, only the soil 

measurements in F and H Area were deemed important for purposes of on-site dose 

reconstruction. 

For purposes of reconstructing potential missed or unmonitored dose, the annual intakes 

provided in Table C-18 should be used only for individuals whose employment records 

indicate that they spent considerable time in either F or H Areas. The 239Pu annual 

intakes should be assumed to have persisted since 1955, and the 238Pu annual intakes 

since 1969. Annual intakes resulting from concentrations of 90Sr and 137Cs should be 

conservatively assumed to have persisted essentially unchanged since 1955.

Section 1.3 above discusses available information regarding chemical forms of 

radioactive materials in use at SRS for major facilities and processes. These chemical 

forms, and the SRS assumptions regarding ICRP 30 inhalation solubility classes, 

provide information to assist the internal dose reconstructor in determining appropriate 

solubility parameters for use with the IMBA program. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 60 of 188

Additionally, Attachment A contains a number of tables providing information related to 

radionuclide mixtures and dosimetric quantities assumed to be applicable for those 

radionuclides as taken from SRS program documentation. It is important to note that 

these dosimetric parameters were based on ICRP 30 recommendations. Internal doses 

reconstructed in support of EEOICPA are derived using more recent ICRP 68 (1994) 

recommended methodology than was in use at SRS. Consequently, inhalation solubility 

classes, dose conversion factors, and annual dose fractions discussed in Section 1.3 

and Attachment A of this document should guide the selection of similar quantities used 

for EEOICPA internal dose reconstructions.

Table 3.1.2-1 Radionuclide Soil Concentrations, pCi g-1, (1 sigma) Dry Weight (0-8 cm 

depth)

Location Sr-90 σ Cs-137 σ Pu-238 σ Pu-239 σ

F-Area 0.0216 0.0182 1.05 0.05 0.403 0.014 0.531 0.014

H-Area 0.0289 0.0183 1.32 0.05 0.0213 0.0016 0.0554 0.0023

Source: CDC 2001, Table 12.2-8, on p. 12.2-8, Arnett 1993

Annual intakes resulting from resuspension of radionuclides in soil in concentrations 

given in Table 3.1.2-1 are presented in Table C-18 and were calculated using the 

following formula:

CA = SD x CS x SF x RF (3-3)

where

CA = Air concentration, pCi m-3 

SD = Soil density = 1600 kg m-3 (PNL-10550, 1995) 

CS = Radionuclide soil concentration = pCi kg-1 

SF = Surface factor = 0.08 m (Arnett 1993)

RF = Resuspension factor = 1.00E-09 m-1 (Till and Meyers, 1983, p. 5-32)

3.2 Occupational External Dose

Workers are subjected to external doses from the ambient radiation levels and from 

noble gases. Prior to 1966, ambient radiation levels were measured with pocket 

ionization chambers and GM detectors. After 1966, the measurements were made with 

ion chambers and thermoluminescent dosimeters (TLDs). Concentrations of the noble 

gas 41Ar were calculated and can be used to determine the external dose due to 

submersion.

3.2.1 Ambient Radiation

The environmental radiological profile has been developed for the Savannah River Site 

for use by dose reconstructors when personal dosimetry or bioassay program 

participation was not required. Site annual environmental reports were reviewed for data 

that would be useful in reconstructing ambient radiation levels and airborne radionuclide 

concentrations. Data in these historical documents (see reference section for citations) 

included ambient ion chamber and thermoluminescent dosimetry (TLD) radiation 

measurements, air sampling results, and residual soil activity. The historical ion 

chamber and TLD results with uncertainty estimates (geometric standard deviations), 

where available, were tabulated and are presented in Table C-19.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 61 of 188

The ambient radiation levels presented in Table C-19 were taken directly from the SRS 

annual and semiannual environmental reports (see reference section) and include 

contributions from the SRS releases and background radiation. Until 1966, ambient 

radiation measurements were performed using either pocket ionization chambers or GM 

detectors, and results were reported in units of millirep (mrep), milliroentgen (mR), or 

millirad (mrad) per 24 hours without any estimates of uncertainty. In 1966, the Savannah 

River Site began performing ambient radiation measurements with TLDs. For purposes 

of this report, mrep, mR, and mrad were assumed to be equivalent to mrem. Although 

ambient radiation results were typically reported in 24 hour increments, measurements 

were usually performed over one-week periods resulting in between 600 and 800 

measurements per year from all locations. The values in Table C-19 are for 2,000 hours 

of exposure per year. 

In 1974, the environmental reports began presenting a quantitative estimate of the 

uncertainty associated with the maximum and minimum radiation measurements. Along 

with the maximum and minimum radiation values, the reports also provided a value for 

two sigma. The 50th percentile values presented in the tables are generally the average 

reported values for a given area. The geometric standard deviation values in Table C-19 

were based on the reported two sigma value. It should also be noted that all of these 

values include the contributions from natural background radiation and nuclear weapons 

fallout. Therefore, use of these data to account for potential missed or unmonitored 

dose should in all cases represent an overestimation of the worker’s actual dose.

3.2.2 Releases of Noble Gases

In addition to radionuclides important to internal dose reconstruction, 41Ar was released 

in relatively large quantities during reactor operation The radionuclide 41Ar is a noble 

(nonreactive) gas with a half-life of 1.8 hours and is produced in the air surrounding the 

reactor. Because this air volume communicated with air in the reactor building, most of 

the 41Ar produced in the reactor was released with ventilation air exiting the 200-foot 

reactor stack. The five reactors were responsible for the vast majority of the 

atmospheric releases of 41Ar. Neutrons interacting with the naturally occurring 40Ar in the 

air surrounding the reactor vessel created 41Ar which quickly mixed with reactor building 

air and was exhausted through the main stack. Once out of the stack, reactor gases 

moved and were dispersed with local air flows and both on-site and off-site individuals 

nearby were potentially exposed to beta and gamma radiation emitted during the 41Ar 

decay. 

The most reliable, reactor-specific, source term estimates, according to CDC (2001, p, 

4.3-1), were provided by Cummins (1991). These source terms were used to calculate 

the 50th percentile annual whole body and organ doses presented in Tables C-20 andC21. Because Cummins (2001) did not include estimates of uncertainty, the 95th

percentile source terms were estimated by assuming that the 95th percentile release was 

simply 25 percent greater than the 50th percentile release.

This potential dose resulting from submersion in the plume of this noble gas is 

considered to be entirely external because the gas is not readily assimilated by the body. 

Using methodology identical to that for the radionuclides important to internal dose, 

calculated air concentrations at the populated areas of the SRS were estimated. These 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 62 of 188

concentrations were then converted to organ and whole body dose using submersion 

dose conversion factors published in Federal Guidance Report No. 12 (EPA 1993). The 

annual doses (assuming 2000 hours per year of exposure) have been tabulated based 

on 50th percentile source terms presented in Tables C-20 and C-21. The geometric 

standard deviations provided in these tables were based on the estimated 95th percentile 

source terms using the relationship provided in Section 3.1.1 above.

For purposes of external dose reconstruction, 41Ar doses to the skin (Table C-20) should 

be assessed as 30% from >15 keV electrons and 70% from >250 keV photons. Doses 

to bone surfaces (Tables C-20 and C-21), should be assumed to be 100% from >250 

keV photons. In addition, 41Ar dose to whole body and other underlying organs (Tables 

C-21) should be assumed to be 100% from >250 keV photons.

3.3 Uncertainty

As discussed in the previous sections, estimates of annual intakes employed 

conservative (i.e., claimant favorable) assumptions. For example, the use of maximum, 

25-year, sector X/Q values to estimate atmospheric radionuclide concentrations. Rollins 

(2003) validated that these calculated air concentrations would likely overestimate the 

actual radionuclide concentrations by comparison to actual measured data taken from 

the SRS annual environmental reports. It should be noted that the estimated annual 

intakes were based on annual average meteorology and that the actual instantaneous 

meteorology on any given day could vary from these averages by factors (plus or minus) 

of several orders of magnitude. However, the intake values given in Attachment C 

should represent a reasonable upper bound of the actual intakes that could have 

occurred.

In instances where more detailed information is known about a particular individual or 

job classification, other modifying factors should be employed. For example, if, for a 

particular job classification, there is reason to believe that the actual ventilation rate for 

the claimant may vary markedly from the average of 2,400 cubic meters per 2,000 hours 

of exposure, the dose reconstructor should use professional judgment to adjust the 

estimated intakes as necessary whether the individual is engaged in light or heavy work. 

The respiration rate for light work is 1.3 m3

/h and 1.7 m3/h for heavy work. In these 

cases, to estimate annual intake, the product of the fractional annual period for each, job 

dependent, level of work and the corresponding ventilation rate, are summed to 

determine the total ventilation volume for the year m3

. The annual intake is then 

determined by multiplying the annual ventilation volume by the annual average 

concentration in Tables C-1 through C-17 for the year and location of interest.

In general, the annual intake values presented in tables C-1 through C-17 reflect 

adjusted source terms and uncertainty analyses provided in CDC (2001). For purposes 

of this document, no other attempt has been made to quantify other uncertainties. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 63 of 188

4.0 OCCUPATIONAL INTERNAL DOSE

Radionuclides of concern listed in the SRS Internal Dosimetry Technical Basis Manual 

(WSRC-IM-90-139, 31 December 2001), chapter 5, include the following elements and 

isotopes:

• tritium (3

H)

• manganese (54Mn)

• cobalt (

60Co)

• zinc (65Zn)

• strontium (90Sr)

• ruthenium (106Ru)

• antimony (125Sb)

• cesium (134Cs, 137Cs)

• cerium (144Ce)

• europium (152Eu, 154Eu)

• thorium (228Th, 232Th)

• uranium (234U, 235U, 236U, 238U, and mixtures)

• neptunium (237Np)

• plutonium (238Pu, 239Pu/240Pu, 241Pu) 

• americium (241Am, 243Am)

• curium (244Cm)

• californium (252Cf, including 248Cm daughter)

These radionuclides were those identified by SRS as the most significant at delivering 

doses or that were effective tracers for significant dose-delivering radionuclides. A 

summary of the data in this section is presented in Attachment D.

4.1 In vitro Minimum Detectable Activities (MDAs), Counting Methods

From WSRC-IM-90-139, the decision level for urinalysis is calculated as:

TREVA

C

D

BACKGROUND

L

* 3

2

65.4

+

= (4-1)

where

DL is in dpm/L

CBACKGROUND = background counts in the region of interest

T = count time

R = recovery

E = average detector efficiency

V = sample volume (usually 1.5 L), and

A = the alpha abundance for the radionuclide in question.

The decision level is the level at which activity is considered present in a sample with a 

95% confidence level.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 64 of 188

Similarly, the MDA is calculated as:

TREVA

C

MDA BACKGROUND *65.4 + 3 = (4-2)

The reporting level is distinguished from the MDA. As described in WSRC-IM-90-139, 

the reporting level is the minimally acceptable decision level; the bioassay laboratory 

was required to achieve decision levels below the reporting level. SRS technical 

documentation provided no quantitative relationship between the reporting level and 

MDA.

4.1.1 In vitro Urine Analysis

Since 1990 excreta analysis records have been maintained in an electronic database. 

The database identifies baseline (new hire), routine, special, and termination 

measurements. Prior to the electronic database, excreta results were kept on employee 

bioassay cards

Current values of MDA for in vitro analyses of routine samples are shown in Table 4.1.1-

1 below, extracted from WSRC-IM-90-139. The MDA values are 2 times the average 

decision level for the analysis. These values are applicable from 1988 through present 

for plutonium, 1990 through present for uranium, and 1994 through present for the 

trivalent actinides americium, curium, and californium. No specific beginning dates were 

found for tritium, 90Sr, or 237Np.

These MDAs apply to both routine urinalyses and to special urinalyses. Since about 

1990, routine analysis results (except tritium) that exceed the decision level are 

recounted for twice the normal count time. The recounted result is placed in the 

database instead of the original count, so the MDA for a recount result will be 

approximately 0.7 times the values shown below. Some samples are “rerun,” meaning a 

second aliquot is processed. In these cases the database will show two results with 

different sample numbers for the same collection date and same sample volume. 

Table 4.1.1-1 Current MDAs for Urinalysis

Radionuclide MDA 

(pCi/L) Radionuclide MDA (pCi/L) Radionuclide MDA (pCi/L)

H-3 20000 U-234 0.032(a) Pu-239 0.022

Sr-90 2.9 U-235 0.036(a) Pu-240 0.022

U-238 0.032(a) Am-241 0.029

Np-237 0.029 Cm-244 0.021

Pu-238 0.032 Cf-252 0.021

(a) Despite the listed MDAs, uranium results below 0.15 dpm/L are not considered “detected.” See the uranium 

discussion in 4.1.2 for details.

Activity fractions for depleted, recycled, and high enriched uranium are listed in the 

glossary section.

Analysis methods for these radionuclides are liquid scintillation counting for tritium, beta 

counting by gas proportional counter for 90Sr, and alpha spectroscopy for the remaining 

radionuclides in table. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 65 of 188

4.1.2 In vitro Methods for Individual Radionuclides

The following section discusses the in vitro methods for specific radionuclides. 

In vitro bioassay for trivalent actinides (americium, curium, and californium)

Records showing urinalysis for trivalent actinides date back at least to the mid 1960s. 

The early reporting levels varied from 3 to 1 dpm/1.5 L. In 1971 the reporting level using 

gross alpha counting on a solid state detector was listed as 0.3 dpm per 1.5 liters. In 

1990, a change in radiochemical processing resulted in an MDA of 0.1 dpm per liter. 

Alpha spectrometry has been used since 1994. For MDAs since 1994, see Table 4.1.1-

1 above. 

Use of a gross alpha count prior to 1994 means that the measurement itself does not 

indicate which radionuclide or combination of radionuclides is being measured. Location 

or other information concerning an intake may allow determination of the principal 

radionuclide. If not, it is claimant-favorable to assume 241Am.

In vitro bioassay for plutonium

From the beginning of the plutonium urinalysis program in 1954 to approximately 1959, 

urine samples were radiochemically processed using bismuth phosphate and lanthanum 

fluoride coprecipitation and electroplated, and activities were determined by gross alpha 

track analysis of exposed NTA emulsions. In 1959 nitric acid/hydrogen peroxide 

dissolution and ion exchange replaced the bismuth phosphate method. This was faster 

and used less volume of urine but had essentially the same MDA. The reporting level 

did not change. Results were recorded as Pu or sometimes as Pu238/Pu239. From 

around 1964 to 1988, counting for gross alpha activity was performed using a solid state 

surface barrier alpha detector. Tri-isooctlyamine (TIOA) liquid extraction replaced the 

ion exchange chemistry in 1966. This method also used direct evaporation on planchets 

instead of electrodeposition. This method also allowed separation of neptunium and 

uranium from the same sample. Sensitivity was stated at 0.1 dpm/1.5L, which was 

consistent with the reporting level already in use. In or about 1981, a new 

coprecipitation technique was introduced for routine samples along with alpha 

spectrometry. Sample-specific determination of plutonium recover by use of a 242Pu 

tracer was also introduced at that time. Results for 238Pu and 239Pu were reported 

separately. The TIOA method with gross alpha counting continued to be used on 

special samples until 1988. An electronic database was introduced in 1990 and results 

were thereafter reported as per liter. Electrodeposition was reinstated in 1994. 

Separation of plutonium+neptunium, actinides, uranium, and strontium from a single 

sample using TEVA and TRU resins began in 2001. Alpha-emitting plutonium and 

neptunium isotopes are electrodeposited and counted by alpha spectrometry on a single 

planchet. Reporting levels for the various radiochemical and counting methods are 

listed in Tables 4.1.1-2 and 4.1.1-3. Note: MDAs are assumed to be more or as 

sensitive as the reporting levels. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 66 of 188

Table 4.1.1-2 Pu-239/240 – in vitro Urinalysis

Time Period Reporting Level(a) Method

1954-1959 0.05 dpm /1.5L

Bismuth phosphate, lanthanum fluoride 

coprecipitation; gross alpha for plutonium; alpha 

track counting

1959-1962 0.05 dpm/1.5L Ion exchange, gross alpha for plutonium, alpha track 

counting (until 1964)

1963-1966 0.1 dpm/1.5L (1963) Solid state alpha counting (1964)

1966-1981 0.1 dpm/1.5L TIOA extraction; gross alpha for plutonium on solid 

state detector 

1981 – 1988 0.07 dpm/1.5L for routine 

samples

Alpha spectroscopy for 239/240Pu (used primarily for 

routine urine samples) 

1988-present 0.06 dpm/L Alpha spectroscopy for 239/240Pu

Table 4.1.1-3 Pu-238 – in vitro Urinalysis

Time Period MDA/Reporting Level –

Urinalysis Urine Analysis Method

1954-1981

Reporting levels same as in Table 

4.1.1-2, 238Pu and 239Pu combined 

into single result 

Methods same as in Table 4.1.1-2 

1981 – 1988 0.05 dpm/1.5 L [reporting level] Alpha spectroscopy for 238Pu (used 

primarily for routine urine samples)

1988-present 0.032 pCi/L (~0.07 dpm/L) [MDA] Alpha spectroscopy for 238Pu

Because plutonium urinalysis results (fecal results as well) prior to 1981 (1988 for 

special samples) measured the total activity from alpha-emitting isotopes whereas 238Pu 

and 239+240Pu results were reported separately from 1981 to the present, the two sets of 

data cannot be entered into the intake code as reported. If, for a specific case, all the 

bioassay data are prior to 1981, the total plutonium alpha data can be entered as 239Pu. 

The calculated intake will have to be partitioned into 238Pu and 239Pu per the mixture 

ratios applicable to the intake (if known). However, it is unlikely that for the older 

intakes, intake-specific isotopic mixture ratios were established so the default mixture 

ratios will have to be used. The same mixture ratio will then be used to determine the 

intake of 241Pu and 241Am. 

Plutonium existed in mixtures of 238Pu, 239Pu, 240Pu, and 241Pu. The activity of 242Pu was 

always insignificant dosimetrically. The relative activities of these isotopes depended on 

the nature of the irradiation of the uranium fuel (referred to a burn-up) and the time 

between the end of irradiation and the intake. Processes at SRS did not perturb the 

relative activities of the isotopes. Generally, SRS created plutonium mixtures ranging 

from about 3% per weight 240Pu to about 12% by weight 240Pu. Generally, the plutonium 

mixtures were blended to produce a final product with about 6% by weight 240Pu, which 

is referred to as a weapons-grade mixture. The 241Pu, a beta-emitter with a 14-yr halflife, undergoes noticeable decay to 241Am during the lifetime of the facilities so that the 

isotopic ratios of contamination will be different than the ratios in newly irradiated fuel (or 

new feed to the Head End Stream). 241Am builds up from near zero at end of irradiation; 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 67 of 188

however, it is removed during production of the plutonium product, then begins to build 

up again as the 241Pu remaining the product decays. So the ratio of 241Pu to, say, 

239+240Pu decreases from time of end of irradiation; whereas the ratio of 241Am to 239+240Pu 

increases from time of last separation of the 241Am from the plutonium. 

If for a given intake the activity of 241Am was established, through chest counting, 

trivalent actinide excreta analysis, or analysis of the source of the contamination, then 

that activity should be used. If unknown, the default ratios provided in Table 4.1.1-4 can 

be used as best fits the nature of the intake. Fresh 12% plutonium mix is the most 

claimant-favorable. 

Table 4.1.1-4 Activity Composition for Reference 6% and 12% 240Pu Mixturesa 

Isotopic 

Component

Reference 6% Pu Mix Reference 12% Pu Mix

Freshb Agedc 5 Years Aged 10 

Years

Fresh Aged 5 Years Aged 10 

Years

Specific Activity in Mixture, Ci/g

Pu-238 8.56E-3 8.23E-3 7.91E-3 1.71E-2 1.64E-2 1.58E-2

Pu-239+240 7.13E-2 7.13E-2 7.13E-2 7.98E-2 7.98E-2 7.98E-2

Pu-241 8.24E-1 6.48E-1 5.09E-1 3.09 2.43 1.91

Am-241 0 5.83E-3 1.04E-2 0 2.19E-2 3.89E-2

Pu alpha 7.99E-2 7.95E-2 7.92E-2 9.69E-2 9.62E-2 9.56E-2

Total alpha 7.99E-2 8.53E-2 8.96E-2 9.69E-2 1.18E-1 1.35E-1

Activity Ratios

Pu239+240:

Am-241

NA 1.22E1 6.87 NA 3.64 2.05

Pu-alpha:

Am-241

NA 1.36E1 7.61 NA 4.39 2.46

Pu-241:

Pu-239+240

1.16E1 9.09 7.14 3.87E1 3.05E1 2.39E1

Pu-241:

Pu-alpha

1.03E1 8.15 6.43 3.18E1 2.53E1 2.00E1

a. From Methods and Models of the Hanford Internal Dosimetry Program (Carbaugh, 2003). Also found in part 

in Guide for Good Practices for Occupational Radiological Protection in Plutonium Facilities (USDOE 1998)

b, Two weeks after end of irradiation

c. Time since separation of the 241Am 

If some of the bioassay data were obtained prior to 1981 and some after, then the data 

will have to be made consistent. This can be done in either of two ways. 1) If all the 

238Pu and 239Pu results are above their respective decision levels, then the 238Pu and 

239Pu results for each sample can be summed to get a total plutonium alpha result. This 

result can be equated to the pre-1981 results and treated in the manner described in the 

preceding paragraph. 2) If some of the 238Pu or 239Pu results are below their decision 

levels, then it may be best to use the isotope that has the most robust data set and 

adjust each of the pre-1981 results to reflect just the portion from that chosen isotope. 

The adjustment will be made using the default isotopic mixture ratios unless the intakespecific ratio is known. For instance a gross alpha result of 2.1 dpm/1.5L would be 

converted to 1.9 dpm/1.5L of 239Pu before being entered into the intake code.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 68 of 188

In vitro bioassay for tritium

From startup until 1958, HTO in urine was analyzed by passing hydrogen evolved from 

the urine sample through an ionization chamber; the reported MDA for this method was 

1 μCi/L. In 1958, liquid scintillation counting was initiated, which is still the method in 

use. A History of Personnel Radiation Dosimetry at the Savannah River Site (WSRCRP-95-234) reports that the MDA consistently improved to the current level of 20,000 

pCi/L (or 0.02 μCi/L). The reporting level remained at the value of 1 μCi L1, and was 

later reduced to 0.5 μCi L-1, on or about 1981, and then to the current value of 0.1 μCi L1

. 

(an exact date was not determined at least by 1989). During the 1980s, although the 

reporting level of 0.5 μCi L-1 was generally used, some results below 0.5 are listed 

directly, e.g. 0.4, 0.3. Considering that the true MDA was probably well below the 

reporting level, these results should be considered as real. 

It should be noted that for current analyses, tritium results of 0.05 μCi L-1 or less are 

reported as “<0.1μCi L-1,” and results between 0.05 μCi L-1 and 0.1 μCi L-1 are reported 

as “0.1 μCi L-1.” Results greater than 0.1 μCi L-1 are reported as measured (to 1 

significant figure).

Tritium analyses are listed as “T” on the employee bioassay cards. Tritium may also be 

listed as “P-10,” especially in the 1950s. Tritium results in the 1990s were listed on the 

same summary form as external dose monitoring results. They are referred to as 

sample results with dates and analysis results, but the word “tritium” or any other 

radionuclide identifier is not mentioned per se. 

For tritium results, the denominator used for reporting purposes has always been per 

liter of urine. (The denominator of 1.5 liters was never used for tritium as it was for other 

radionuclides.)

Only tritium bioassay results greater than 5 μCi L-1 were evaluated prior to computer 

evaluation of data.

In vitro bioassay for uranium

A variety of methods have been used historically to analyze uranium at SRS. These 

methods and the associated detection capabilities are summarized in Table 4.1.1-5.

Activity fractions for depleted, recycled, and highly enriched uranium are listed in the 

glossary section. SRS technical documentation indicates that for earlier monitoring 

periods, the designations “enriched” and “depleted” analysis for uranium referred to 

analysis performed by alpha counting or chemical measurement, respectively, and was 

not necessarily indicative of the degree of uranium enrichment. EU was the code used 

on employee bioassay cards for the gross alpha count method, and U was used to 

designate the fluorophotometric method. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 69 of 188

Table 4.1.1-5 Uranium Urinalysis

Time Period Uranium 

mixture Reporting Level Urine analysis method

Startup to mid1960sEnriched 

uranium 0.15 dpm/1.5 L Gross alpha for uranium, alpha track counting

Mid-1960s to 

1982

Enriched 

uranium 1 dpm/1.5 L Gross alpha for uranium on solid state detector

1954 - 1982 Depleted 

uranium 1-5 μg/L Fluorophotometric analysis

1982 - 1986

U-235 0.14 ng Delayed neutron analysis 

(DNA)

U-nat 1 μg/L DNA analysis for U-235

Enriched 

uranium 1 dpm/L DNA analysis for U-235

1986-1990 Enriched 

uranium 1 dpm/1.5 L Gross alpha for uranium on 

solid state detector

1986-1994 Depleted 

uranium

[not listed in WSRC-RP95-234]Kinetic phosphorimetry 

analysis (KPA)

1990 – present 

(enriched 

uranium), 1994-

present 

(depleted 

uranium)

DU use U-238 and isotopic 

ratios from glossary

Alpha spectroscopy for 

specific uranium isotopes

RU use U-238 and isotopic 

ratios from glossary

HEU use U-235 and isotopic 

ratios from glossary 

U-234 0.032 pCi/L

U-235 0.036 pCi/L

U-238 0.032 pCi/L

Because of excretion of uranium from natural sources, the MDAs for natural or depleted 

uranium listed above should not be used to determine occupational intakes. Based on a 

probit analysis of 1728 worker samples in 1993-1994 (WSRC-IM-90-139, rev. 8), an 

upper limit at the 99% confidence level for analytical noise and presumably some effect 

of natural background was set at 0.15 dpm/L. The SRS policy is that a sample result 

must exceed both its decision level and 0.15 dpm/L to be considered a detection. 

Lacking other information, i.e., similar studies at different times throughout the history of 

the SRS, it is reasonable to apply the 0.15 dpm/L criterion to all historical samples. The 

background excretion in terms of μg/L was not established, but almost certainly was 

below the 1-μg/L reporting level.

If isotopic data are available, they can also be used to determine if a uranium urinalysis 

result represents an occupational intake. The following rules apply:

• If 235U or 236U are detected, then 234U must also be detected.

• If both 234U and 238U are detected, the uranium should be considered from 

occupational sources if the 234U/238U ratio is outside of a range from 1 to 3. 

• If 238U is detected, the uranium should be considered from an occupational source if 

the ratio of the 234U critical level to the 238U result is less than 1.

• If 234U is detected, the uranium should be considered from an occupational source if 

the ratio of the 234U result to the 238U critical level is greater than 3. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 70 of 188

In vitro analysis for radiostrontium

The start date for urinalysis for radioisotopes of strontium was not specifically 

determined, other than “late 1950s.” From start until 1969 strontium was separated by 

alkaline earth phosphate coprecipitation followed by beta counting on a GM counter. 

This analysis was also called fission product analysis (see discussion below). Both 89Sr 

and 90Sr would have been counted (and potentially radioisotopes of cerium and 

promethium), but it is claimant-favorable to assume the result is all 90Sr. Yttrium-91, 

grown in from 91Sr, might also be present at intake if the source of intake was recent 

fission products (up to about a year post irradiation). If the intake prior to 1969 occurred 

in reactor buildings or the 221-F canyon facility and an intake of 90Sr is indicated, then an 

equal intake of 91Y should be assumed. This is a claimant-favorable assumption 

because after about a year post irradiation, 91Y activity would be negligible. The 

reporting level is assumed to be the same as for the beta component of the fission 

product analyses. 

From 1969 to 2001 strontium was analyzed by liquid ion-exchange that separates the 

yttrium progeny, followed by beta proportional counting. Yttrium-91 would be included 

as a possible interference but 89Sr would not. For 90Sr intakes in 1969 to 1989 in reactor 

buildings, an intake of 89Sr equal to the intake of 90Sr should be assumed. 

Since 2001, 90Sr has been separated as part of the use of TRU and TEVA resins for 

plutonium, uranium, actinides, and strontium. 

In vitro fission product analysis

Prior to whole body counts, intakes of gamma-emitting fission products were monitored 

by urinalysis. The record is not clear exactly what procedure was used; both gammaspectrometry analyses and gross beta counting are mentioned as being possible. Nor 

has it been determined if any separation chemistry was done prior to counting, though 

the reporting levels that were used seem to imply that 40K was removed or reduced prior 

to counting. These analyses were coded FP, FPIA, or IA under element on the 

employee bioassay cards. The IA code can have two meanings depending on the 

context. An IA code with a result prior to 1960 most likely refers to the fission product 

analysis; an IA code with no result means inconclusive analysis, that is, the lab was 

unable to produce a valid result. The employee bioassay cards show less-than values 

that varied over the years, presumably indicating different reporting levels and probably 

indicating a switch from beta counting to photon counting; for instance, 1955-1960: <30 

d/m/750 ml; 1960-65: <100 d/m/1.5 L; 1966-1971; <1 nCi/1.5L. Some records in the 

1960-65 period show two results for the same sample: one for beta (<100 dpm/1.5L) and 

one for gamma (<500 dpm/1.5L). This would imply they did a gamma scan, probably on 

the whole sample before the chemistry. The records also show some fission product 

urinalyses as late as 1984 with the result listed as <1. Presumably the units are 

nCi/1.5L. 

A procedure for fission product urinalysis at Hanford, dated May 1950, has been 

discovered (Thorburn, 1950). It is likely that the Savannah River procedure was similar; 

at least through 1960. The procedure separated and counted radionuclides of strontium, 

yttrium, cerium, and promethium. It did not account for radionuclides of ruthenium or 

cesium. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 71 of 188

To interpret results from the fission product urinalysis, it is noted that the strontium, 

barium, europium, zirconium, and niobium radionuclides concentrate primarily in the 

bone, with 90Sr providing the largest dose rate. Cerium, lanthanum, and promethium 

concentrate primarily in the liver, with some concentration in the bone, with 144Ce 

providing the largest dose rate. Cesium and ruthenium are assumed to be uniformly 

distributed in the whole body. 

Hence, based on what little is known, fission product urinalysis results should be 

interpreted as follows:

Note: If the claimant had a whole body count during any year and had a detectable 

fission product urinalysis result, the intakes of 137Cs, 106Ru, or any other gamma-emitting 

fission products should be determined from the whole body count. 

• 100% Type F 90Sr if the claimant’s disease is related to bone or bone marrow plus an 

equal intake of 137Cs (Type F) and 3 fold intake of 106Ru (Type F). 

• 100% Type W 144Ce if the claimant’s disease if related to the liver plus an equal 

intake of 106Ru and a 20% intake of 137Cs (Type F). 

• For any other disease location, the fission product analysis should be assumed Type 

F 90Sr to be an indicator of a 137Cs intake (although it probably did not measure 137Cs 

directly). Hence, calculate an intake of 90Sr from the fission product results and add 

an equal intake of 137Cs and 3 fold intake of 106Ru. 

In vitro analysis for 237Np

The Savannah River Technical Basis Manual, Rev. 8, says that neptunium analyses 

started in the 1950s, perhaps 1959, using coprecipitation with calcium-magnesium 

ammonium phosphate and ion exchange, followed by autoradiography on NTA emulsion 

for 10,000 minutes. The MDA was not given but was mostly likely similar to the MDA for 

plutonium alpha autoradiography method used at the same time, i.e., about 0.035 

dpm/L. The results were recorded as S.P. (special product), and the reporting level was 

0.05 dpm/1.5 L, same as plutonium. Analysis by TIOA extraction and electrodeposition 

occurred at the same time as for plutonium analyses, about 1964, which was a faster 

method but the reporting level stayed the same. In 1993-1996, samples were analyzed 

by an offsite, commercial laboratory using extraction chromatography resin. In 1996 

analysis returned to SRS and alpha spectrometry was introduced. Since 2001, TEVA 

and TRU resins are used to separate neptunium from actinides and uranium, and alpha 

spectrometry has continued. The MDA is still listed as 0.035 dpm/L; however, from 

personal communication with the SRS lead internal dosimetrist, an MDA of 0.4 dpm/L 

was considered more applicable in practice. If results are reported below 0.4 dpm/L, 

they can be used as given. 

4.1.3 Correcting for Urinalysis Volume

At SRS, routine urine samples are considered partial-day samples. Urinalysis results 

are recorded as per 1.5L or per L. The code used to determine intake from bioassay 

data requires the period of time represented by the sample and the result of the sample 

in activity per day. If the volume of the sample is known, the result can be adjusted to 

activity per sample. Then if the collection time of the sample is known, the activity can 

be adjusted to activity per day by ratioing the collection hours to 24 hours. (This is often 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 72 of 188

the case for special samples collected since about 1990.) If either the collection period 

or the volume is not known, then a result in activity per L or per 1.5 L should be 

normalized to activity per day using the reference volume of 1.4 L per day. 

4.1.4 Fecal Sample Analysis

SRS technical documentation indicates that fecal sample analysis has been performed 

periodically since the 1950s. These samples were always for special, rather than 

routine, sampling purposes, and were always performed in conjunction with urine 

bioassay sampling. Fecal results will be found on separate reports, not routine employee 

bioassay cards.

Fecal samples are screened by photon counting with the following typical MDAs: 7 pCi 

(

241Am), 300 pCi (238Pu), and 600 pCi (weapons grade plutonium). However, if results 

could not be quantified by photon counting, that is, if nothing was detected, then 

samples may have been analyzed by normal separation chemistry and alpha 

spectrometry to improve the detection capability. This protocol applies mostly to 

samples from 1994 to present, and probably was not applied to samples prior to 1994. 

Because only an aliquot of the fecal sample is used for wet radiochemistry, the MDAs 

are typically 1.5 times the MDAs for routine urine samples. 

4.2 In vivo MDAs, Counting Methods, and Reporting Practices

Whole body counting has been performed since 1960 when the SRS Whole Body 

Counting facility was completed. Chest counting was initiated in the early 1970s using 

phoswich detectors. In vivo count records are not kept in a central electronic database; 

each count is maintained as a single hard-copy record. A whole body count record and 

a chest count record on the same day are listed on a single hard-copy record. In vivo

bioassay methods and detection capabilities are reviewed below.

4.2.1 Whole Body Counting

Prior to whole body counting, fission products intakes were monitored using urinalysis.

Whole body counting began in approximately 1960, using a 4” high by 8” thick diameter 

NaI detector. (Detection energy range was 100 keV to 2000 keV.) The monitored 

individual sat in a reclining chair positioned in an arc around the detector; this was 

referred to as the “40-cm arc geometry” in bioassay monitoring reports. Bioassay via this 

method was not used for plutonium and americium due to their low energy emissions. It 

is reported this detector was in service until September 1995. Reported MDAs for 

various radionuclides are shown in Tables 4.2.1-1 and 4.2.1-2. MDAs and reporting 

levels were the same
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 73 of 188

Table 4.2.1-1 Whole Body Counting

Period Nuclide MDA, nCi MDA method

~1960 - ? 

Ce-144 29

~3X standard deviation 

of expected count rate 

in energy region

U(a) 62

I-131 1.4

Ru-106 6.1

Cs-137 1.0

Zr/Nb-95 2.2

Zn-65 5.1

Ba/La-140 9.3

(a) Listed as U, but based on measurement of U-235. A result greater than the MDA would 

certainly be a false positive unless it was associated with a major intake. Other counts and 

urinalyses would be expected.

In the early 1970s, whole body counting using an array of NaI detectors was 

implemented. (WSRC-IM-90-139 indicates five 4”x4” detectors were used, while WSRCRP-95-234 says that four 4”x5” detectors were used.) Additionally, in the mid-1980s, 

whole body counting using stand-up geometry and large (4”x4”x16” or 5”x3”x16”) NaI 

detectors was implemented. MDAs for these counting systems were generated 

individually for counts by processing software. Many forms for reporting the results of 

whole body counts were used throughout the years. Some had the MDAs listed on the 

form; others just reported “detected” radionuclides. If MDAs are not shown, use the 

ones in Table 4.2.1-1.

Note: Some of the forms use the abbreviation nc for nanocurie.

Table 4.2.1-2 Current MDAs for Whole Body Counting

Radionuclide MDA (nCi) Radionuclide MDA (nCi) Radionuclide MDA (nCi)

Mn-54 3.4 Sb-125 14 Eu-152 18

Co-60 2.9 Cs-134 3.8 Eu-154 8.4

Zn-65 6.1 Cs-137 4.1 U-235 14

Ru-106 36 Ce-144 69 Np-237 14

For special investigational counts, the MDAs are stated to be approximately 0.71 times 

these values for whole body counting.

Many SRS workers had/have burdens of 137Cs from non-occupational sources, e.g., 

fallout and consumption of venison. The SRS presently uses a value of 20 nCi of 137Cs 

in a whole body count for consumers of venison as the decision level to follow up or not. 

If the record clearly indicates that the worker was a consumer of meat from wild game 

harvested on the mid-to-southern Atlantic seaboard extending into the Appalachian 

Mountains, 137Cs results in whole body counts can be disregarded unless there is also 

an indication of intake of other fission/activation products or 90Sr in a urine sample. 

Otherwise, it is claimant-favorable to assume the 137Cs resulted from an occupational 

intake. 

4.2.2 Chest Counting

Chest counting was initially performed using a 5” diameter thin NaI detector from 1966 to 

around 1971. No MDA information could be located in SRS technical documentation for 

this method of analysis. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 74 of 188

In 1971, a phoswich detection system was implemented that used a NaI/CsI sandwiched 

detector. This system was modified to a dual phoswich system in 1972, which remained 

in use until around 1987 or 1988. MDAs were listed on results forms that varied from 

count to count but were in the neighborhood of 0.2 nCi for 241Am, 5 nCi for 238Pu, 10 nCi 

for 239Pu, and 5 nCi for 244Cm; however, based on operating experience over the years, 

the values for all but the 241Am seem overly optimistic due to noise in the 1-20 keV part 

of the spectrum. The method of accounting for chest wall thickness may also be a 

factor in the difference between stated MDAs and present experience. However, 

phoswich systems had better low-energy background and better detection efficiency 

than germanium systems. Phoswich systems had poor resolution and could not 

discriminate between 238Pu, 239Pu, or 244Cm. 

In general chest counting MDAs should not be used to determine potential undetected 

intakes of plutonium if urinalysis data are available.

The current six detector solid state planar germanium array was implemented in 1987 or 

1988.

MDAs for later periods have been generated individually for counts by processing 

software, with typical values shown in Table 4.2.2-1, based on a chest wall thickness of 

2.5 cm. However, MDAs are often listed for each count as part of the measurement 

results and the count-specific MDAs should be used when available. The uncertainty in 

the MDA calculation has been estimated, according to WSRC-IM-90-139, rev. 8, as 

±46%. The confidence level was not stated but would appear to be one standard 

deviation. However, uncertainty in chest wall thickness can have a large impact on MDA 

determination. Using weight and height measurements on 1100 male workers at 

Hanford in 2002-2003, and using the chest wall thickness formula in the SRS technical 

basis manual, the 50th percentile chest wall thickness was 3.7 cm and the 95th percentile 

chest wall thickness was 5.9 cm. The MDAs for radionuclides with low-energy photons 

affected by variations in chest wall thickness for the 50th and 95th percentile male 

workers are shown in Table 4.2.2-2. 

. 

Table 4.2.2-1 Current MDAs for Chest Counting (2.5 cm chest wall thickness)

Radionuclide MDA (nCi) Radionuclide MDA (nCi) Radionuclide MDA (nCi)

Ce-144 0.31 U-235 0.10 6% Pu 96

Eu-152 0.056 U-236 89 12% Pu 70

Th-228 3.2 U-238 1.1 Am-241 0.10

Th-232 31 Np-237 0.31 Am-243 0.12

DU 1.2 Pu-238 58 Cm-242 28

RU 8.3 Pu-239 130 Cm-244 37

HEU 5.2 Pu-240 47 Cf-252 32

U-234 30 3% Pu 110
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 75 of 188

Table 4.2.2-2 Current MDAs for Chest Counting 50th Percentile and 95th Percentile Male 

Workers

Radionuclide MDA (nCi)

50th

MDA (nCi)

95th

Am-241 0.13 0.20

Cm-244 78 310

Pu-238 130 580

Pu-239 290 1300

Pu-240 110 470

Cf-252 85 500

For special investigational chest counts, the MDAs are stated to be approximately 0.77 

times these values.

For 241Am results associated with an intake of 6% to 12% 240Pu mixture, the dose 

reconstructor can assume that the plutonium isotopes behave the same in the lung as 

the 241Am. This assumption does not hold for systemic americium, however. 

4.3 Personal Air Sampling Data as Applicable

The objective of breathing zone air monitoring is to measure the concentration of

radioactive material in the air that a worker breathes for the purpose of calculating 

exposure [in Derived Air Concentration-hours (DAC-h)]. Intakes calculated from air 

monitoring were assigned only if bioassay data are unavailable or inadequate. Breathing 

zone air monitoring must sample air that is representative of the air in the worker's 

breathing zone. This was accomplished by placing the sampler as close as possible to 

the worker's nose and mouth. Fixed breathing zone air samplers are typically placed at 

face height in front of hoods and glove boxes. Portable breathing zone air samplers, 

often referred to as personal air samplers (PAS), are attached to the worker's clothing in 

the neck region. PAS have been used when it was not practical to use regional air 

samplers (RAS) (such as work outside of buildings), inside plastic suits used in highly 

contaminated environments, and for operational studies.

Air sample records were considered workplace monitoring records as opposed to intake 

monitoring records and hence were not maintained in the same records collection as 

bioassay records. Air sample records were associated with each facility separately and 

were stored as facility records. Boxes containing workplace monitoring records can be 

retrieved from storage at the federal repository; however, locating a specific set of air 

sample records would be very labor consuming and should be considered only as a last 

resort. Correlation between air sample concentrations in given rooms or work locations 

and a specific person would be difficult. 

4.4 Interferences, Uncertainty, and Use of Reporting Levels

4.4.1 Contamination of Samples

Excreta samples were collected in the workplace. Because the levels of activity that are 

significant in excreta samples, especially urine samples, were generally below 

detectability on workplace personnel detectors, contamination of samples from the 

worker’s hands or clothing is a possibility. Hanford found a decrease in detectable 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 76 of 188

plutonium bioassay results after switching to home collection. Laboratory contamination 

and mix-up of samples in the laboratory are also a possibility, although laboratory 

Quality Control procedures and performance of test samples was designed to minimize 

this source of contamination. 

It is likely that a contaminated sample will show up as an obvious outlier in the dataset 

for a given worker. If the dataset shows an unusually high urinalysis result for a 

radionuclide other than tritium, and if follow-up samples were collected that were not 

consistent with the high result, then the high result may be considered an outlier. 

However, if the result is not obviously an outlier, then it is claimant-favorable to assume 

the result is real. 

For in vivo measurements, contamination can occur as external to the body or, in the 

case of chest counting, as external to the lung. If a follow-up in vivo count is obtained 

the same day or within a few days that shows a dramatic decrease in activity or no 

detectable activity, then external contamination can be assumed. Radon progeny and 

medical diagnostic or therapeutic procedures involving radionuclides can cause 

interferences to in vivo measurements, especially for NaI detectors. However, unless 

the count was invalidated or noted as being influenced by such interferences, the results 

should be used as recorded. 

4.4.2 Uncertainty

Uncertainties for the bioassay measurements were not stated in the records. For results 

near or at the reporting levels, the assumption provided in the Internal Dose 

Reconstruction Implementation Guide (NIOSH 2002b) should be used, namely the 

standard deviation is 0.3 times the MDA or reporting level. For results greater than 3 

times the MDA or reporting level, the standard deviation can be assumed to be 0.1 times 

the result, based on Currie’s quantification level (Currie 1968). If actual standard 

deviations or other indications of error are reported with a bioassay measurement result, 

the reported value should be used. 

4.4.3 Use of Reporting Levels and Less-Than Values

Because plutonium stays in the body for a very long time, and because the urine excreta 

curve (activity per day excreta versus days after intake) has a small slope beyond the 

first year, an intake of plutonium that might have been missed in the 1950s or 60s 

because of poor detection capability or other reasons can still be confirmed or otherwise 

by urinalysis obtained years later. Additionally, low-level chronic or frequent-intermittent 

intakes not originally detectable can, if continued for many years, will lead to increasing 

rates of excreta over time. This condition has implications for the determination of 

potential missed intakes.

Special consideration for plutonium intakes 

Table 4.4.3-1 provides the maximum potentially missed chronic intakes of 239Pu based 

on a urinalysis MDA of 0.1 dpm/day (0.07 dpm/L) for various periods of chronic intake 

and two commonly used particle size distributions. The urine sample is obtained on the 

last day of the period (under Duration of Intake). This approach provides the maximum 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 77 of 188

potentially missed chronic intake regardless if there were missed samples or samples 

with high MDAs during the period. Once the 239Pu intake is obtained, standard isotopic 

ratios for (e.g., 6% 240Pu or 12% 240Pu) can be applied to get the total intakes. 

Table 4.4.3-1 239Pu Chronic Inhalation Intake Assessment Based on MDA Urinalysis on 

Last Day of the Period

Solubility: Type M Type S

Particle Size: 5 5

Analytical MDA: 0.1 0.1

Duration of Intake

Daily Intake 

Rate Cumulative Intake

Daily Intake 

Rate Cumulative Intake

(years) (days) (dpm/d) (dpm) (nCi) (dpm/d) (dpm) (nCi)

5 1825 15.86 2.89E+04 13.0 293.4 5.35E+05 241

10 3650 11.37 4.15E+04 18.7 164.3 6.00E+05 270

15 5475 9.432 5.16E+04 23.3 121.6 6.66E+05 300

20 7300 8.243 6.02E+04 27.1 99.17 7.24E+05 326

25 9125 7.410 6.76E+04 30.5 84.98 7.75E+05 349

30 10950 6.785 7.43E+04 33.5 75.09 8.22E+05 370

35 12775 6.293 8.04E+04 36.2 67.76 8.66E+05 390

40 14600 5.894 8.61E+04 38.8 62.11 9.07E+05 408

45 16425 5.564 9.14E+04 41.2 57.59 9.46E+05 426

50 18250 5.284 9.64E+04 43.4 53.90 9.84E+05 443

Solubility: Type M Type S

Particle Size: 1 1

Analytical MDA: 0.1 0.1

Duration of Intake

Daily Intake 

Rate Cumulative Intake

Daily Intake 

Rate Cumulative Intake

(years) (days) (dpm/d) (dpm) (nCi) (dpm/d) (dpm) (nCi)

5 1825 10.00 2.01E+04 9.04 164.2 3.00E+05 135

10 3650 7.847 2.86E+04 12.9 90.0 3.29E+05 148

15 5475 6.500 3.56E+04 16.0 66.24 3.63E+05 163

20 7300 5.677 4.14E+04 18.7 53.90 3.94E+05 177

25 9125 5.102 4.66E+04 21.0 46.14 4.21E+05 190

30 10950 4.670 5.11E+04 23.0 40.75 4.46E+05 201

35 12775 4.331 5.53E+04 24.9 36.77 4.70E+05 212

40 14600 4.056 5.92E+04 26.7 33.70 4.92E+05 222

45 16425 3.828 6.29E+04 28.3 31.26 5.13E+05 231

50 18250 3.635 6.63E+04 29.9 29.26 5.34E+05 241
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 78 of 188

5.0 OCCUPATIONAL EXTERNAL DOSIMETRY

5.1 Introduction

SRS operations involved several processes of the nuclear weapons development cycle 

(DOE 1997) and represented a significant role within the US nuclear weapons program. 

These processes include nuclear fuel fabrication; nuclear reactor operations; 

radiochemical separations; radionuclide production; recycling uranium; refining, finishing 

and storing plutonium; and handling the associated radioactive waste. 

SRS workers, especially those employed during the peak production decades of the 

1950s and 1960s, have been exposed to radiation types and energies associated with 

the respective nuclear weapon development processes. SRS utilized facility and 

individual worker monitoring methods to measure and control radiation exposure to 

workers. There are numerous SRS records concerning facility monitoring, safety 

evaluations, investigations, etc. However, it is time consuming to locate and evaluate 

these records for SRS facilities and processes since 1951. Evaluations of the extensive 

scope of facility, process, and worker information relevant to an individual worker’s 

potential dose, many years or even decades after employment are difficult. Records of 

radiation dose to individual workers from personnel dosimeters worn by the worker and 

co-workers are available for SRS operations beginning in 1951. Dose from these 

dosimeters was recorded at the time of measurement, routinely reviewed by SRS 

operations and radiation safety staff for compliance with radiation control limits, and 

routinely available to individual workers. The NIOSH External Dosimetry Implementation 

Guide (NIOSH 2002a) has identified these records to represent the highest quality 

record to retrospectively assess dose. Information in this section pertains to analyzing 

these records and does not address parameters regarding skin and testicular and/or 

breast radiation dose that may result from acute beta (electron) radiation exposure in 

generally non-routine workplace exposure profiles. 

Radiation dosimetry practices were initially based on experience gained during several 

decades of radium and x-ray medical diagnostic and therapy applications. These 

methods were generally well advanced at the start of the Manhattan Engineering District 

(MED) program to develop nuclear weapons beginning about 1940. The primary 

difficulties encountered by MED operations to measure worker dose to external radiation 

involved:

• Large quantities of high level radioactivity

• Mixed radiation fields involving beta, photon (gamma and x-ray), and neutron 

radiation with low, intermediate, and high energies.

• Neutron radiation.

5.2 Basis of Comparison 

Historically, since the initiation of the MED in the early 1940s, various radiation dose concepts 

and quantities have been used to measure and record occupational dose. The basis of 

comparison for reconstruction of dose, as described in “External Dose Reconstruction 

Implementation Guideline” (NIOSH 2002a), is the Personal Dose Equivalent, Hp(d), where d 

identifies the depth (in mm) and represents the point of reference for dose in tissue. For weakly 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 79 of 188

penetrating radiation of significance to skin dose, d = 0.07 mm and is noted as Hp(0.07). For 

penetrating radiation of significance to “whole body” dose, d = 10 mm and is noted as Hp(10). 

Both Hp(0.07) and Hp(10) are the radiation quantities recommended for use as the operational 

quantity to be recorded for radiological protection purposes by the International Commission on 

Radiological Units and Measurements (ICRU 1993). In addition, Hp(0.07) and Hp(10) are the 

radiation quantities used in the DOE Laboratory Accreditation Program (DOELAP) used to 

accredit DOE personnel dosimetry systems since the 1980s (DOE 1986). The International 

Agency for Research on Cancer (IARC) Three Country Combined Study (Fix et al 1997) and 

IARC Collaborative Study (Thierry-Chef et al 2002) selected Hp(10) as the quantity to assess 

error in historical recorded “whole body” dose for workers in the IARC nuclear worker 

epidemiologic studies. The basis for comparison for neutron radiation is more complicated 

since historically the calibration of dosimeters to measure neutron dose was based on different 

dose quantities such as First Collision Dose, Multiple Collision Dose, Dose Equivalent Index, 

etc. The numerical difference in using these dose quantities compared to the Hp(10) dose used 

in current DOELAP performance testing could be evaluated using the relative values of the dose 

conversion factors for the respective dose quantities in conjunction with characteristics of the 

respective SRS neutron dosimeter response characteristics and workplace radiation fields. 

5.3 Dose Reconstruction Parameters

Examinations of the beta, photon (x-ray, gamma ray), and neutron radiation types 

energies and geometries of exposure, and the characteristics of the respective SRS 

dosimeter response are crucial to the assessment of bias and uncertainty of the original 

recorded dose in relation to the radiation quantity Hp(10). The bias and uncertainty for 

current dosimetry systems is typically well documented for Hp(10) (SRS 1993). Often 

the performance of current dosimeters can be compared with performance 

characteristics of historical dosimetry systems in the same, or highly similar, facilities or 

workplaces. In addition, current performance testing techniques can be applied to earlier 

dosimetry systems to achieve a consistent evaluation of all dosimetry systems. 

Dosimeter response characteristics for radiation types and energies in the workplace are 

crucial to the overall analysis of error in recorded dose.

Overall, accuracy and precision of the original recorded individual worker doses and 

their comparability to be considered in using NIOSH (2002) guidelines depend on (Fix et 

al 1997):

• Administrative practices adopted by facilities to calculate and record personnel 

dose based on technical, administrative, and statutory compliance considerations.

• Dosimetry technology includes the physical capabilities of the dosimetry system, 

such as the response to different types and energies of radiation, in particular in 

mixed radiation fields.

• Calibration of the respective monitoring systems and similarity of the methods of 

calibration to sources of exposure in the workplace.

• Workplace radiation fields that may include mixed types of radiation, variations in

exposure geometries, environmental conditions.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 80 of 188

An evaluation of the original recorded doses based on these parameters is expected to 

provide the best estimate of Hp(0.07), as necessary, and Hp(10) for individual workers 

with the least relative overall uncertainty. 

5.3.1 SRS Historical Administrative Practices

Historically, SRS had an extensive radiation safety monitoring program to measure 

radiation exposure in the workplace using portable radiation instruments, contamination 

surveys, zone controls, and personnel dosimeters. The results from the personnel 

dosimeters were used to measure and record dose from external radiation exposure to 

SRS workers throughout the history of SRS operations (Taylor et al 1995). These 

dosimeters include one or more of the following:

• Personnel Whole Body (WB) beta/photon dosimeters 

• Pocket Ionization Chamber (PIC) dosimeters 

• Personnel extremity dosimeters

• Personnel neutron dosimeters. 

SRS began operations in 1951 using dosimeter and processing technical support 

provided by the Oak Ridge National Laboratory (ORNL). A summary of SRS personnel 

beta/photon and neutron (respectively, dosimeter type, exchange, Minimum Detection 

Level (MDL) and potential missed annual dose) is shown in Table 5.3.1-1. ORNL, then 

the Clinton Laboratory, had implemented their dosimetry methods based on the 

personnel beta/photon dosimeter design developed at the Metallurgical Laboratory at the 

University of Chicago (Pardue et al 1944). ORNL provided SRS with beta/photon film 

dosimeters and neutron nuclear track, type A (NTA) emulsion dosimeters at the 

beginning of SRS operations until SRS implemented their in-house personnel dosimetry 

capabilities about one year later. The precise detection levels shown in Table 5.3.1-1 

are difficult to estimate, particularly for the older systems. The current SRS commercial 

TLD system minimum detection levels (MDLs) are identified in SRS external dosimetry 

technical basis documentation (SRS 1993) based on a DOE Laboratory Accreditation 

Program (DOELAP) laboratory testing protocol (DOE 1986). During earlier years, the 

MDLs are subject to additional uncertainty since factors involving the radiation field and 

film type, as well as the processing, developing and reading system cannot be tested. 

The film dosimeter MDLs shown in Table 5.3.1-1 were estimated based on information 

from NIOSH (1993), NRC (1989), and Wilson et al (1990). The TLD MDLs were 

obtained from Taylor et al (1995) and SRS (1993). 

Dose reconstruction parameters concerning SRS administrative practices significant to 

dose reconstruction involve issues as follows:

• Policies to assign dosimeters to workers

• Policies to exchange dosimeters

• Policies to record notional dose (i.e., some identified value for lower dosed workers 

often based on a small fraction of the regulatory limit) 

• Policies to estimate dose for missing or damaged dosimeters

• Policies to replace any destroyed or missing records

• Policies to evaluate and record dose for incidents
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 81 of 188

• Policies to obtain and record occupational dose to workers for other employer 

exposure.

Table 5.3.1-1 SRS Dosimeter Type, Period of Use, Exchange Frequency, MDL and 

Potential Annual Missed Dose

Dosimeter Period of Use Exchange 

Frequency

Laboratory 

MDL 

(mSv)(a)

Max. Annual 

Missed Dose

(mSv)(b)

Beta/Photon Dosimeters

ORNL TwoElement film 

Dosimeter

Prior to 3/52 Weekly (n = 52) 0.4 20.8

SRS Two 

Element film 

Dosimeter

After 3/52 through 

9/30/1957

Weekly (n = 52) 0.4 20.8

Beginning 10/1/1957 

through 11/8/1959

Biweekly (n=26) 0.4 10.4 

SRS MultiElement Film 

Dosimeter

Beginning 11/9/1959 

through 12/31/1964

Biweekly (n=26) 0.4 10.4

Beginning 1/1/1965

through 12/31/1965

4-week (n=13) 0.4 5.2

Beginning 1/1/1966 

through 3/31/1970

Monthly (n=12) 0.4 4.8

SRS TLD Beginning 4/1/1970 

through 6/30/1983

Monthly (n=12) 0.15 1.8

Panasonic TLD Beginning 7/1/1983 to 

2003 (ongoing)

Monthly (n=12) 0.05 0.6

Neutron Dosimeters

ORNL NTA Prior to 8/3/1953 Weekly (n = 52) ~0.5 13(c)

SRS NTA Beginning 8/3/1953 

through 7/13/1960

Weekly (n = 52) ~0.5 13(c)

SRS NTA Beginning 7/14/1960 

through 12/31/1970

Biweekly (n=26) ~0.5 6.5(c)

SRS Hoy TLND Beginning 1/1/1971 

through 12/31/1994

Monthly (n=12) 0.2 1.2

Panasonic TLND Beginning 1/1/1995 to 

2003 (ongoing)

Monthly (n=12) 0.15 0.9

a. Estimated film dosimeter detection levels based on NIOSH (1993), NRC (1989), and Wilson et 

al (1990). TLD detection levels from Taylor et al (1995) and SRS (1993). 

b. Maximum annual missed dose (NIOSH 2002)

c. The potential annual missed dose based on laboratory irradiations is not applicable to 

workplace missed neutron dose. 

SRS policies appear to have been in place for all of these parameters. SRS routine 

practices appear to have required assigning dosimeters to all workers who entered a 

controlled radiation area. Dosimeters were exchanged on a routine schedule. All 

dosimeters were processed and the measured results recorded and used to estimate 

dose. There appears to be no use of recorded notional doses, although there are issues 

of recording dose for low-dose dosimeters (See section on “missed dose”). Figure 5.3.1-1 

illustrates the number of SRS monitored workers and the collective “whole body” dose of 

record for SRS workers since 1951 (Taylor et al 1995). Administrative practices are 

generally available in the SRS procedures manual (DuPont 1954, 1957, 1961a, 1961b), 

and detailed information for each worker in the respective SRS exposure history file 

documentation. Based on the SRS documentation and the reasonable trend in these 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 82 of 188

figures, there does not appear to be any significant administrative practice that would 

affect the integrity of the recorded dose of record for SRS workers with the exception 

that SRS hard-copy radiation records must be consulted to separate the photon, neutron 

and tritium dose for workers who terminated prior to 1979. Apparently, only the total 

whole body dose (i.e., photon + neutron + tritium) is available from the second quarter of 

1963 through December 31, 1972. The missing dose components of the total whole 

body dose could be estimated from prior year and subsequent year dose results 

according to the methods described by Watson et al (1994) based on examination of 

continuity in the worker’s job and work activities. 

0

5000

10000

15000

20000

25000

1940 1950 1960 1970 1980 1990 2000

Year

Normalized Values

Monitored Workers

Whole Body Dose

Figure 5.3.1-1 SRS Monitored Workers and Collective Dose

5.3.2 SRS Dosimetry Technology

SRS dosimetry methods evolved during the years as improved technology was 

developed and the complex radiation fields were better understood. The adequacy of 

the respective dosimetry methods to accurately measure radiation dose is determined 

from the radiation type, energy, exposure geometry, etc., as described in later sections. 

The dosimeter exchange frequency was gradually lengthened, generally corresponding 

to the time period of the regulatory dose controls. At the beginning of SRS operations, a 

weekly dose control of 3 mSv was in effect. This was changed to an annual limit of 50 

mSv in the latter 1950s. Table 5.3.2-1 is a summary of the major historical events in the 

SRS personnel dosimetry program. 

Operations at SRS occurred about 7 to 8 years after the MED nuclear weapons facilities 

began operations at Oak Ridge National Laboratory (ORNL), Los Alamos National 

Laboratory (LANL), and Hanford. These AEC sites implemented, in the beginning, 

essentially the same beta/photon and neutron dosimeter methods. There were 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 83 of 188

intercomparisons of dose interpretations among these AEC sites through the years as 

reported in Wilson et al (1990). 

Table 5.3.2-1 SRS Historical Dosimetry Events (Taylor et al 1995)

Date Description

1951 Pocket Ionization Chambers (PICs) were used to measure SRS worker exposure. 

PICs were used in addition to film dosimeters.

11/12/1951

ORNL two-element (i.e., open window and 1 mm silver filter) beta/photon 

dosimeters issued to SRS workers. Processing was provided by ORNL. Routine 

dosimeter exchange period was weekly.

1951

Nuclear Track Emulsion (NTA) dosimeters were issued to selected SRS workers to 

measure neutron radiation. Capability was provided by ORNL. NTA film was 

placed, facing the worker’s body, in two-element beta/photon dosimeter holders. 

NTA film exchange was also weekly.

3/3/1952 SRS began onsite dosimetry program with the same two-element film dosimeter 

design as ORNL. 

3/1953 SRS began processing film using the ORNL film badge dosimeter. The results 

were reported in mrem/week, manually recorded on individual worker file folders. 

1953 Extremity film dosimeter use and processing began at SRS.

8/3/1953 SRS assumed processing of NTA films.

4/1954 Indium foil placed in dosimeter holder to respond to neutron radiation.

10/1/57 SRS beta/photon dosimeter exchange changed to biweekly.

11/9/1959 SRS designed multi-element film dosimeter use was implemented. This dosimeter 

design permitted analysis of beta, gamma and x-ray exposure to personnel. 

7/14/1960 NTA exchange changed to biweekly.

1963 PICs were replaced with self-reading dosimeters to provide a real-time indication of 

exposure.

1965 Four week beta/photon dosimeter exchange implemented.

1966 Monthly beta/photon dosimeter exchange implemented. 

4/1/1970 SRS TLD replaced film as the principal means of measuring beta/photon dose. The 

minimum recorded dose was about 0.15 mSv.

1/1/1971 SRS Hoy Thermoluminescent Neutron Dosimeter (TLND) implemented.

1980 Extremity dose routinely measured with TLD chip.

7/1/1982 Commercial Panasonic beta/photon TLDs implemented. 

mid-1980s

SRS implemented changes such as on-phantom calibration and 137Cs calibration 

source to participate in new DOE Laboratory Accreditation Program (DOELAP) 

implemented to performance testing. The overall change in recorded dose with 

new practices, as described in Taylor et al (1995), was determined to be an 

equivalent 11.9% increase compared to earlier dose estimates.

1989

SRS TLD is DOELAP accredited. Lower limit of detection, based on DOELAP 

protocol, was 0.2 and 0.05 mSv, respectively, for the shallow and deep dose 

components. 

1/1/1995

Panasonic neutron dosimeter replaced SRS TLND. Routine dosimeter exchange 

for the Panasonic beta/photon dosimeter is quarterly and for the Panasonic TLND, 

monthly.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 84 of 188

5.3.2.1 Beta/Photon Dosimeters

In the early 1950s, SRS used personnel dosimeter services from ORNL (Taylor et al 

1995). ORNL had implemented in the mid-1940s the beta/photon film dosimeter design, 

with perhaps minor differences, that was developed originally at the Metallurgical 

Laboratory at the University of Chicago (Pardue et al 1944). Within about one year, 

SRS implemented their own in-house dosimetry capabilities. SRS followed a process of 

dosimetry research and development leading to gradual upgrades in dosimetry 

capabilities for the complex radiation fields encountered in SRS reactor, reprocessing 

and plutonium handling facilities (Taylor et al 1995). This evolution in dosimetry 

capabilities was followed by other DOE sites leading to site-specific multi-element film 

and thermoluminescent dosimetry systems. SRS dosimeters always included a 

comparatively “open window” to measure significant beta radiation and to distinguish the 

dosimeter response between beta and photon radiation. The photon energy response 

characteristics of the SRS beta/photon dosimeters are illustrated in Figure 5.3.2.1-1 

based on the essentially identical two-element film dosimeter used at SRS, Hanford and 

ORNL (and likely other MED sites) designed at the University of Chicago (Pardue et al 

1944). Figure 5.3.2.1-1 also illustrates the Hp(10) response. The improved energy 

response of multi-element film and thermoluminescent dosimeters is shown in Figure 

5.3.2.1-1 using available Hanford published information (Wilson et al 1990). The general 

trend in the energy response curves for these dosimeters, relative to Hp(10), provides

insight into the expected performance of these dosimeters in SRS workplace radiation 

fields. The original two-element dosimeter shows a significant under-response to photon 

energies less-than about 100 keV. Evaluations of the two-element film dosimeter that 

was used at SRS, ORNL, LANL, and Hanford sites have been performed by the 

International Agency for Research on Cancer (IARC) Collaborative Study researchers 

(Thierry-Chef et al 2002) and by Fix et al (1994) to several photon energies. IARC 

conducted a dosimeter intercomparison study to higher-energy (i.e., >100 keV) photons 

of ten commonly used historical dosimetry systems throughout the world. Three of the 

dosimeter designs were from the US. These included the two-element dosimeter used 

at Hanford (identified as US-2), multi-element film dosimeter used at Hanford (identified 

as US-8), and the currently used SRS Panasonic dosimeter system (identified as US22). The IARC Study considered that exposure to dosimeters worn by workers could be 

characterized as a combination of anterior-posterior (A-P), rotational, and isotropic 

irradiation geometries. Dosimeter response to selected photon energies was measured 

using two phantoms. These phantoms were used to simulate the effect of the workers 

body on the measured dosimeter response. The first phantom is the International 

Standards Organization water filled slab phantom with polymethyl methacrylate walls 

measuring 30 cm wide by 30 cm tall and 15 cm in depth that is widely used for dosimeter 

calibration and performance testing. The second phantom used was an 

anthropomorphic Alderson Rando Phantom. This phantom is constructed from a natural 

human skeleton cast inside material that has a tissue equivalent response. The results 

of this testing, for US dosimeters only, are presented in Table 5.3.2.1-1. As noted in 

Table 5.3.2.1-1, the two-element dosimeter design significantly overestimated Hp(10) for 

all irradiation geometries and increasingly for the lower energy photon energy at 118 

keV. Similar anthropomorphic phantom laboratory irradiations using anterior-posterior 

and rotational geometries to several photon energies have been done at Hanford 

(Wilson et al 1990) with similar results. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 85 of 188

0.1

1

10

100

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

Photon Energy, MeV

Relative Respons

Hp(10)

Two-Element Film OW

Two-Element Film Shield

Multi-Element Film

TLD

Figure 5.3.2.1-1 Estimated SRS Dosimeter Photon Response Characteristics

To evaluate the dosimeter response to lower-energy (i.e., < 100 keV) photons that are 

significant in plutonium facilities, Hanford conducted intercomparison testing of all 

Hanford historical dosimeter film designs (Wilson et al 1990) using A-P irradiations only. 

Although there are differences in films and particularly the filters used in the multielement dosimeters, good comparison in energy response for SRS and Hanford 

dosimeters is expected. The results of this testing is summarized in Table 5.3.2.1-2. 

The dosimeter results for energies greater than 100 keV, are consistent with the IARC 

results showing an over-estimate of Hp(10) for the two-element dosimeter. For energies 

less than 100 keV, the two-element dosimeter underestimates the photon dose. This 

under-response is also illustrated in the original University of Chicago two-element 

dosimeter energy response curve (Pardue et al 1944). 

Table 5.3.2.1-1 IARC Testing Results for US beta/photon Dosimeters.

Geometry Phantom

118 keV 208 keV 662 keV

Mean(a) SD/ Mean Mean(a) SD/ Mean Mean(a) SD/ Mean

US-2 (Two-Element Film Dosimeter)

A-P Slab 3.0 2.1 1.3 1 1.0 0.8

A-P Anthropomorphic 3.0 4.2 1.2 1.9 1.0 1.8

Rotational Anthropomorphic 2.2 2 1.4 3 1.2 3.2

Isotropic Anthropomorphic 1.5 4.4 1.1 1.6 1.0 2.7

US-8 (Multi-Element Film Dosimeter)

A-P Slab 1.0 1.5 1.0 0.8 0.8 1.7

A-P Anthropomorphic 0.8 9.5 0.9 6 0.8 1.8

Rotational Anthropomorphic 1.2 1.9 1.2 17 1.1 1.8

Isotropic Anthropomorphic 1.0 3 1.2 9 1.0 2.3

US-22 (Multi-Element Thermoluminescent Dosimeter)

A-P Slab 0.9 4.4 0.9 3.9 0.9 3.5

A-P Anthropomorphic 0.8 3.1 0.9 2.1 0.9 3.9

Rotational Anthropomorphic 1.1 3.1 1.2 1.5 1.0 4.1

Isotropic Anthropomorphic 0.9 0.3 1.0 2.5 0.9 1.6

(a) Ratio of recorded dose to Hp(10).
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 86 of 188

SRS adopted a conservative practice1 historically to calculate the deep dose for selected 

workers in plutonium facilities in specific instances when gonadal dose was considered 

significant. This practice involved adding the open window (i.e., skin dose), based on a 

16 keV k-fluorescent calibration, to the deep dose as measured from the film response 

behind the 1 mm silver filter. Figure 5.2.3.1-1 illustrates the potential significance of this 

correction to the routinely recorded dose. Based on this practice, routine SRS deep dose 

estimates are judged to be reasonably accurate compared to Hp(10) for all photon 

energies. SRS used uranium to routinely calibrate the dosimeter open window dose 

response.”

Table 5.3.2.1-2 Testing Results for Hanford Two-Element and Multi-Element Film 

Dosimeters for Energy and Angular Response(a,b) 

Anterior-Posterior (AP) Exposure Rotational Exposure

 Film Dosimeters Film Dosimeters

Beam

(Energy, 

keV)

Two 

Element

1944-56

MultiElement,

1957-71

TLD,

1972-

Present

Two 

Element, 

1944-56

MultiElement,

1957-71

TLD,

1972-93

16(b) 0.1 0.9

59(b) 0.5 1.1

M150(70) 0.7 0.70 0.95 1.31 1.31 1.77

H150(120) 1.6 0.64 0.87 3.00 1.20 1.64

137Cs(662) 1.0 1.0 1.0 1.46 1.46 1.46

(a) Divide recorded dose by table value to estimate Hp(10).

(b) Based on Wilson et al (1990).

5.3.2.2 Neutron Dosimeters 

Two general types of SRS neutron dosimeters have been used and these differ 

dramatically in their response to neutron radiation. 

NTA - The nuclear track emulsion, type A, (NTA) neutron dosimeter consisted of the 

NTA film enclosed in the same holder as used for SRS beta/photon dosimeters until 

November 1959 when a NTA specific dosimeter holder was implemented (Taylor et al 

1995). 

TLD - The SRS TLD neutron (TLND) dosimeter was implemented on 1 January 1971. 

The TLND is a sphere design, using two polyethylene spheres covered on the curved 

surface with cadmium. The dosimeter is worn using a belt to minimize distance from the 

worker’s body (Hoy 1972). SRS implemented a commercial Panasonic neutron TLD 

system on January 1, 1995.

The AEC held a series of Personnel Neutron Dosimetry Workshops to address problems 

being experienced by the respective AEC sites concerning accurate measurement of 

neutron dose. The first workshop was held 23–24 September 1969 (Vallario et al 1969) 

with the stated concern: “...for intermediate energy ......neutron sources, NTA personnel 

neutron dosimeters cannot be effectively used. This leaves a gap in the personnel 

dosimetry program which at many installations may be quite serious.” The workshops 

 

1 NIOSH Procedures manual.... And discussions during June 19, 2003 teleconference among NIOSH, 

NIOSH contractor and SRS technical staff.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 87 of 188

were generally limited to representatives from sites with active personnel neutron 

dosimetry programs and continued for many years. The 11th DOE Workshop on 

Personnel Neutron Dosimetry was held in 1991 (Rabovsky et al 1991). The significance 

of the underestimated neutron dose became evident with the studies leading to 

implementation of the thermoluminescent dosimeters (TLDs) and using simultaneous 

NTA and TLD (neutron) measurements of recorded worker dose. Hanford (Nichols et al 

1972) conducted field studies with workers wearing both types of dosimeters. 

Response characteristics of the SRS NTA type dosimeter and the TLND neutron 

dosimeter in comparison with other DOE neutron dosimeters are presented in 

Brackenbush et al (1980). In general, the response of the NTA dosimeter decreases 

with decreasing neutron energy that are greater than a threshold energy estimated by 

SRS to be about 500 keV (Taylor et al 1995) and the TLD response increases with 

decreasing neutron energy. The response of both dosimeters is highly dependent upon 

the neutron energy spectra. Both dosimeter types require matching the laboratory 

calibration neutron spectra with the workplace spectra for accurate results. Results 

reported by Hoy during the first AEC Neutron Dosimetry Workshop in 1969 indicated that 

simultaneous laboratory2 measurements of the dose equivalent with NTA was about 

one-half to one-fourth that compared with other methods including the TLND (Vallario et 

al 1969). 

5.3.3 Calibration

Potential error in recorded dose is dependent on the dosimetry technology response 

characteristics to each radiation type, energy, and geometry; the methodology used to 

calibrate the dosimetry system; and the extent of similarity between the radiation fields 

used for calibration and that present in the workplace. The potential error is much 

greater for dosimeters with significant variations in response such as film dosimeters for 

low energy photon radiation and both the NTA and TLND for neutron radiation. 

5.3.3.1 SRS Beta/Photon Dosimeters

SRS dosimeters were originally calibrated using primarily uranium and 226Ra using in-air 

(i.e., no phantom) exposures to selected levels. K-fluorescent x-rays were used to 

develop dosimeter response characteristics for the lower energy photon fields in 

plutonium facilities. This practice is similar to other AEC sites. Taylor et al (1995) 

describes adjustments to SRS recorded dose to estimate Hp(10) based on SRS 

preparations for DOELAP performance testing in the mid-1980s. At that time, it was 

concluded that:

• Prior to 1 January 1986 the recorded dose of record (i.e., photon) dose should be 

multiplied by a factor of 1.119 (11.9%). 

• Prior 1 January 1987 recorded dose of record (i.e., photon) should be multiplied by a 

factor of 1.039 (3.9%).

These changes in the recorded dose should be made to arrive at an assured claimantfavorable treatment. Common sources of laboratory bias are shown in Table 5.3.3.1-1 

 

2 Based on June 19, 2003 teleconference between NIOSH, NIOSH project contractors and SRS technical staff.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 88 of 188

for personnel beta/photon dosimeter calibration based on comparison of the recorded 

dose with Hp(10). 

5.3.3.2 SRS Neutron Dosimeters

Historical aspects of the calibration of SRS NTA and TLNDs are described by Taylor et 

al (1995) and in articles by Hoy that are included in several reports during Atomic Energy 

Commission workshops on personal neutron dosimetry such as the first during 23-24 

September 1969 (Vallario et al 1969) and second during 8-9 July 1971 (Vallario et al 

1971). Table 5.3.3.2-1 describes the common sources of laboratory bias for personnel 

neutron dosimeter calibration based on expected comparison of the recorded dose with 

Hp(10). 

Table 5.3.3.1-1 Common Sources of Laboratory Bias for Beta/photon Dosimeter 

Calibration Parameters(a)

Parameter Historical Description Anticipated Laboratory Bias(b

In-Air 

Calibration

In the 1980s, SRS began exposing 

calibration dosimeters on phantom 

(used to simulate worker’s body). 

As such, previous calibrations do 

not include response from radiation 

backscatter response.

Recorded dose of record too high. 

Backscatter radiation from worker’s body is 

highly dependent upon dosimeter design.

Radiation 

Quantity Prior to the 1980s, SRS dosimeter 

systems were typically calibrated 

to photon beam measured as 

exposure.

For higher energy 226Ra and 137Cs gamma 

radiation used to calibrate dosimeters this 

caused only a slight (about 3%) underresponse in recorded dose. 

Tissue depth of 

dose

Historically, SRS used selected 

depths of 1 or 2 cm to estimate the 

deep dose. 

The numerical effect of this for photon radiation 

is comparatively low. SRS dosimeter designs 

had filtration density thickness of about 1,000 

mg/cm2 that would relate to the 1 cm depth in 

tissue. 

Angular 

Response

SRS dosimeter system is 

calibrated using anterior-posterior 

(A-P) laboratory irradiations.

Recorded dose of record likely too low.

Dosimeter response is lower at non A-P 

angles. Effect is highly dependent upon 

radiation type and energy. 

Environmental 

Stability

SRS film dosimeter and TLD 

systems are subject to signal fade 

with time, heat, humidity, light, etc.

Recorded dose of record likely too low

however this depends strongly upon when 

calibration dosimeters are irradiated. Mid-cycle 

calibration minimizes effects. 

a. Judgment based on SRS dosimeter response characteristics.

b. Recorded dose compared to Hp(10).
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 89 of 188

Table 5.3.3.2-1 Common Sources of Laboratory Bias for Neutron Dosimeter Calibration 

Parameters(a)

Parameter Historical Description Anticipated Laboratory Bias(b)

Source Energy 

SpectraIn the 1980s, SRS began exposing calibration dosimeters on phantom 

(used to simulate worker’s body). 

As such, previous calibrations do 

not include response from radiation 

backscatter response. The beam 

was also degraded in an attempt to 

better represent the workplace.

The delivered dose was uncertain as noted in 

Brackenbush et al (1987). (see workplace 

radiation fields)

Radiation 

Quantity

Neutron dose quantities used to 

calibrate SRS neutron dosimeter 

systems have varied historically. 

For many years the neutron dose 

equivalent index was used.

The effects of the respective dose quantities 

used to calibrate SRS neutron dosimeters is 

uncertain and could be evaluated in 

comparison to the Hp(10) dose used in 

DOELAP performance testing.

Angular 

Response

SRS dosimeters calibrated using 

anterior-posterior (A-P) laboratory 

irradiations.

Recorded dose of record likely too low since 

dosimeter response is lower at angles other 

than A-P. Effect is highly dependent upon 

energy. 

Environmental 

Stability

SRS NTA film dosimeter and TLD 

systems are subject to signal fade 

with time, heat, humidity, light, etc.

Recorded dose of record likely too low

because of fading however this effect 

depends strongly upon when calibration 

dosimeters are irradiated. 

a. Judgment based on SRS dosimeter response characteristics.

b. Recorded dose compared to Hp(10). 

5.3.4 Workplace Radiation Fields

SRS operations are characterized by significant complex beta, photon and neutron 

radiation fields in SRS reactor, irradiated fuel processing, plutonium handling, neutron 

source production, and radioactive waste facilities. Descriptions of processes and 

primary nuclides are presented in the respective SRS External Dosimetry Technical 

Basis documents (WSRC 2003).

5.3.4.1 SRS Workplace Beta/Photon Dosimeter Response

Essentially all SRS radiological work areas involved beta/photon radiation covering a 

wide range of energies characteristic of the radionuclides being handled within the 

respective facilities and processes. Radiation beta/photon fields characteristic of SRS 

facilities can be generally defined based on historical information as presented in Table 

5.3.4.1-1. The expected performance of SRS dosimeters to these fields has been 

estimated in this TBD using the extensive evaluations of the very similar Hanford 

dosimeters (Wilson et al 1990, Fix et al 1994, 1997), the IARC intercomparison study 

results (Thierry-chef et al 2002) and available internal3 SRS documentation

 

3 For example, letter, dated September 3, 1968 from W.R. Tyson, to J.D. Ellett entitled, “Personnel 

Radiation Exposures – Building 235-P.”
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 90 of 188

Table 5.3.4.1-1 Selection of Beta and Photon Radiation Energies and Percentages for SRS 

Facilities 

Process/

Buildings Description

Operations Radiation 

Type

Energy 

Begin End Selection Percent

Fuel 

Fabrication

Produced reactor fuel and target 

assemblies. 1953 1989

Beta

Photon

> 15 keV

30 – 250 keV

100%

100%

300M

Uranium fuel fabrication 1953 1989

Reactors

During Operation: Highly dispersed fields of higher-energy 

photon radiation fields from fission process, activation/ fission 

product nuclides.. Potential for significant airborne nuclides and 

there may be significant higher-energy beta radiation. 

Beta

Photon

> 15 keV

30 – 250 keV

> 250 keV

100%

50%

50%

While not in Operation: Highly dispersed fields of higher-energy 

photon radiation fields from activation/fission product nuclides.. 

There may be significant higher-energy beta radiation during 

maintenance work resulting from fission products.

105C

Heavy Water Moderated Reactors

3/1955 6/1985

105K 10/1954 1988

1992 7/1992

105L 7/1954 1968

1985 6/1988

105P 2/1954 8/1988

105R 12/1953 6/1964

Processing

Plants

Radiochemical Operations: Highly dispersed fields of higherenergy photon radiation fields from activation/fission product 

nuclides dominant to most exposure profiles. Potential for higherenergy beta radiation during sampling and, maintenance work 

resulting from fission products. 

Beta

Photon

> 15 keV

30 – 250 keV

> 250 keV

100%

50%

50% 200F Radiochemical processing plant 1954

200H Radiochemical processing plant 1955

Plutonium 

Production

Plutonium Component Production: Plutonium is machined into 

weapon components using a glove box assembly process with 

predominant close anterior exposure to workers. Radiation 

characteristics in this area involve significant lower-energy 

photons and lower-energy neutron radiation.

Photon < 30 keV

30 – 250 keV

25%

75%

Plutonium Storage: Radiation characteristics in this area 

generally involve dispersed lower-energy neutron radiation and 

scattered photons to include 60 keV Am-241 gamma ray.

221F, B-line Plutonium Finishing Process 1954

221H, B-line Plutonium Finishing Process 1955

772F Production Control Laboratory 1955

235F Plutonium Fuel Facility 1955

736A Calibration Facility 1953

Radionuclide

Production

Radiation characteristics are highly dependent upon the 

radionuclide being produced. SRS has produced several 

radionuclides including tritium, 238Pu, 239Pu, americium, curium, 

californium, etc. 

(Specific to 

radio-nuclide)

Photon, 

default

30 – 250 keV

> 250 keV

 

50%

50%

773A Savannah River Laboratory 1953

Waste 

Handling

Radiation characteristics are highly dependent upon the source of 

waste but typically fission product nuclides (90Sr/90Y, 137Cs) are 

dominant. 

Beta

Photon

> 15 keV

30 – 250 keV

> 250 keV

100%

50%

50%

200F General waste handling, storage 

tanks, etc.

1953 Ongoing

200H General waste handling, storage 

tanks, etc

1953 Ongoing
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 91 of 188

Common beta/photon personnel dosimeter parameters important to Hp(10) performance 

in the workplace are summarized in Table 5.3.4.1-2. Based on the collective 

information, SRS dosimeters are expected to reasonably measure the Hp(10) dose 

under all SRS workplace radiation fields. The SRS historical practice, for the twoelement film dosimeter in plutonium facilities characterized by predominant photon 

energies <100 keV, to calculate the total whole body deep dose by summing the shallow 

dose from the open window film response based on a 16 keV fluorescent x-ray 

calibration and the deep dose from the 1 mm silver filtered film response based on a 

226Ra calibration, will result in a over-estimate of the actual Hp(10) dose. In addition, 

laboratory irradiations of the two-element dosimeter have shown an over-response of the 

actual Hp(10) dose by about a factor of 2 to photons greater than 100 keV. A claimantfavorable approach is proposed to ignore this over-response because of the complexity 

of workplace photon energies and exposure geometries that tend to result in an underestimate of the Hp(10) dose. The respective SRS dosimeters have filtration of 

approximately 1000 mg/cm2 (i.e., nearly equivalent to 1 cm depth in tissue) for those 

regions of the dosimeter used to measure the whole body dose. The response to beta 

radiation in SRS workplaces is limited because beta radiation cannot penetrate the 

filtration. As such, a reasonable estimate of deep dose, compared to Hp(10), is 

expected for SRS beta/photon workplace radiation. 

Table 5.3.4.1-2 Common Workplace Beta/Photon Dosimeter Hp(10) Performance(a)

Parameter Description Workplace Bias(b)

Exposure 

Geometry

SRS dosimeter system 

calibrated using anteriorposterior (A-P) laboratory 

irradiations.

Recorded dose of record likely too low since dosimeter 

response is lower at angles other than A-P. Effect is 

highly dependent upon radiation type and energy. 

Energy 

Response

SRS film “deep dose” 

dosimeter response to 

photon radiation <100keV 

is too low and >100keV is 

too high.

Bias in recorded dose depends upon the photon 

spectra in the workplace. SRS practice to include 

the shallow dose based on a 16-keV calibration to the 

deep dose for selected workers in SRS plutonium 

facilities where gonadal dose was significant eliminated 

this common source of potential under-response. 

Reasonable estimate of Hp(10) dose is expected.

Mixed Fields

SRS dosimeters respond 

to beta and photon 

radiation.

Filtration of about 1,000 mg/cm2 over dosimeter 

component used to calculate deep dose will minimize 

dosimeter response to beta radiation. Reasonable 

estimate of Hp(10) dose is expected.

Missed Dose

Doses less than Minimum 

Detection Level recorded 

as zero dose.

Recorded dose of record likely too low and the relative 

impact is significant primarily in earlier years with 

frequent dosimeter exchange and film dosimeters with 

higher MDLs. 

Environmental 

Effects

Workplace heat, humidity, 

etc., fades dosimeter 

signal. 

Recorded dose of record likely too low.

a. Judgment based on SRS dosimeter response characteristics and workplace radiation fields.

b. Recorded dose compared to Hp(10).
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 92 of 188

5.3.4.2 SRS Workplace Neutron Dosimeter Response

There are four main areas at SRS where there is a potential for neutron exposure: the 

plutonium facilities in the 200 area; the Calibration Facility (736-A) and the Cf-252 facility 

(773-A) in 700 area; the reactor areas (100) during operation; and the 321 Building (PuAl alloys) in the 300 Area. Each of these facilities has different neutron exposure 

spectrum. These areas also have demonstrated different neutron to photon ratios. The 

following sections discuss the different energy spectra and neutron to photon ratios by 

facility. 

5.3.4.2.1 Plutonium Facilities

The majority of the neutron exposure at the Savannah River Site (SRS) is associated 

with the separation, finishing, and storage of plutonium. These operations are 

performed primarily on the 200F B-line (221-F), the 200H HB-line (221-H), the 

Production Control Laboratory (772-F), and at the PuFF Facility (235-F). Neutron dose 

is also associated with process control analysis and storage of the product. The 

plutonium is introduced into the SRS plutonium finishing processes as plutonium 

fluoride. The plutonium fluoride is converted into plutonium oxide and made into 

buttons. Plutonium fluoride produces an (α,n) reaction resulting in a significant source of 

neutrons. Neutrons are also emitted spontaneously from 238Pu and 240Pu, which 

compose a large activity fraction of the plutonium buttons. Plutonium-238 is the primary 

isotope used for the production of heat sources (Bebbington, 1990). 

5.3.4.2.1.1 Neutron Energy Spectra

Brackenbush et al (1987) measured neutron spectra at several locations across the 

Savannah River Site. Figure 5.3.4.2-1 provides the neutron spectra measured at glove 

boxes on the FB and HB lines respectively. It should be noted that the data in 

Brackenbush et al. (1987) were depicted as dose equivalent rates, however for simplicity 

of calculation, a one hour exposure was assumed in order to use dose equivalent. 

Neutron Energy (MeV)

10-7 10-6 10-5 10-4 10-3 10-2 10-1 100 101 102

Dose Equivalent (mrem) 10-3

10-2

10-1

100

101

102

0.355 mrem 0.168 mrem

36.0 mrem

0.557 mrem

FB Line Glovebox

Neutron Energy (MeV)

10-7 10-6 10-5 10-4 10-3 10-2 10-1 100 101 102

Dose Equivalent (mrem) 10-4

10-3

10-2

10-1

100

101

0.304 mrem 0.118 mrem

2.60 mrem 2.83 mrem

HB Line Glove Box 

Figure 5.3.4.2-1 Neutron Energy Spectra Recorded at SRS Plutonium Facilities with Neutron 

Energy Groups Overlay.

Although somewhat different neutron spectra were recorded for the different exposure 

areas (HB Line, FB Line, and storage facility), similarities were observed in the general 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 93 of 188

shape of the spectra and the energy of the flux peak at approximately 1-2 MeV. The HB 

glovebox spectra in Figure 5.3.4.2-1 demonstrated a higher peak energy which is 

possibly the result of more spontaneous fission from Pu-238 but other spectra from the 

HB line indicated a spectra with neutron energies between the FB Glove box line spectra 

and the HB Glovebox spectra depicted. 

The fraction of the total dose in each neutron energy group was determined by 

subdividing the neutron spectra into the four lower neutron energy groups discussed in 

the External Dose Reconstruction Implementation Guideline (NIOSH 2002) (See 

overlays in Figure 5.3.4.2-1. The highest neutron energy group (>20 MeV) was not used 

since operations at SRS did not produce neutrons of this energy. The dose for each 

neutron energy group was calculated by multiplying the neutron flux (Ø) provided in 

Brackenbush et al (1987) by the corresponding flux to dose-rate conversion factors 

(DCF) found in NCRP 38. The neutron doses in each NCRP 38 energy interval are then 

summed to develop the four neutron group doses. The dose fraction (Df) for each 

neutron energy group (n) was then calculated by dividing the neutron group dose by the 

total dose (DT) (Eqn 5.1). 

T

i

ii

nf D

E DCF

ED

∑

=

)(

)(

φ

 (5.1)

where:

Ø(Ei

) = Neutron flux of the ith energy bin

DCFi = NCRP 38 (1971) flux to dose-rate conversion factor for the ith energy bin

DT = Total dose

Tables 5.3.4.2.1.1-1 and 5.3.4.2.1.1-2 provide the neutron dose fractions by energy 

group using the data measured by Brackenbush et al (1987).

Table 5.3.4.2.1.1-1 Measured Dose Fractions at Different Locations in H Area

Neutron 

Energy Group

HB-Line 

Glovebox 

HB-Line in Front 

of Hood

Through 

Glove Port 

(extremity)

Default

< 10 keV 0.041 0.003 0.028

10 – 100 keV 0.026 0.004 0.002

0.1 – 2 MeV 0.438 0.584 0.563

2 – 20 MeV 0.495 0.409 0.407

Claimant Favorable Dose Fractions

0.1 – 2 MeV 0.5 0.6 0.6 0.6

2 – 20 MeV 0.5 0.4 0.4 0.4

 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 94 of 188

Table 5.3.4.2.1.1-2 Measured Dose Fractions at Different Locations in F Area

Neutron 

Energy Group

FB-Line 

Glovebox Storage Area Default

< 10 keV 0.009 0.029

10 – 100 keV 0.007 0.017

0.1 – 2 MeV 0.966 0.816

2 – 20 MeV 0.018 0.138

Claimant Favorable Dose Fractions

0.1 – 2 MeV 1.0 0.86 1.0

2 – 20 MeV 0.14

 

Although Brackenbush et al. (1987) measured some dose from lower (<10 keV) and 

intermediate (10-100 keV) energy neutrons, the contribution to the total dose was less 

than 5%, respectively. The Radiation Effectiveness Factors used in the Interactive 

RadioEpidemiological Program (IREP) to calculate the Probability of Causation (PC) for 

these two neutron energy groups are less than the fission spectrum energy group (0.1 – 

2 MeV) (Kocher et al., 2002). As a result combining the lower and intermediate energy 

neutron groups into the fission spectrum group is a reasonable and claimant favorable 

simplification of the dose calculation. In both table 5.3.4.2.1.1-1 and 5.3.4.2.1.1-2, 

claimant favorable default values are also provided when there is insufficient work 

history information to determine the type of operation the energy employee conducted. 

5.3.4.2.1.2 Neutron to Photon Ratio

Trends in the annual SRS and Hanford neutron collective dose (Taylor et al 1995, 

Buschbom and Gilbert 1993, respectively), normalized to the annual plutonium 

production (DOE 1996), are illustrated in Figure 5.3.4.2-2. It is evident in this figure that 

the collective neutron dose was under-recorded prior to implementation on 1 January 

1971 of the SRS TLND and 1 January 1972 for the Hanford TLD. The extent of the 

under-estimate is difficult to estimate. SRS and Hanford showed a significant increase 

in the ratio of the annual collective neutron dose to the annual plutonium production 

when the TLD neutron dosimeters were implemented. The continued increase in the 

ratio shown in Figure 5.3.4.2-1 may be attributable to the increasingly lower handling of 

plutonium as done during production compared to increasing work activities in later 

years to store plutonium with the associated high scatter of neutrons from shielding. The 

TLND is well known to over-respond to lower energy neutrons (Scherpelz et al 2002). 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 95 of 188

Year

1940 1950 1960 1970 1980 1990

Collective Neutron Dose / Pu Production

0

50

100

150

200

250

300

350

400

450

SRS 

Hanford 

TNLD use began

Hanford

SRS

Figure 5.3.4.2-2 Trends in SRS and Hanford Collective Neutron Dose Normalized to 

Plutonium Production.

Subsequent to the implementation of the Hanford TLD on 1 January 1972, AEC 

headquarters staff conducted a detailed review of recorded neutron dose for Hanford 

personnel using a committee of technical experts from Hanford, SRS, and other AEC 

facilities (Biles 1972). Central to this investigation was the selection of 18 long-term 

Hanford workers for detailed evaluation. Fix et al (1996) conducted a later analysis of 

this information using additional dosimetry data recorded after 1972. Three distinct 

periods of dose recording from 1950 through the present are identified in Table 

5.3.4.2.1.2-1 corresponding to the respective Hanford two-element, multi-element and 

thermoluminescent dosimeters that parallel similar dosimeters implemented at SRS. 

Table 5.3.4.2.1.2-1 Recording Periods for Selected Hanford Plutonium Workers 

Category Period Description

A 1950-56

This period of record involved use of the original Hanford two-element 

dosimeter for nonpenetrating (shallow) and penetrating (deep) dose 

components and NTA film for neutron radiation. 

B 1957-71

This period of record involved use of the Hanford multi-element film 

dosimeter for nonpenetrating, x-ray and penetrating dose components, 

and NTA film for neutron radiation. 

C 1972-95 This period of record involved use of the Hanford TLD for beta, photon, 

and neutron dose components.

All of these workers had the Hanford plutonium facility as their primary work area at least 

during the 1970s. Nine of the 18 workers have dose histories that extend from 1950 or 

earlier through 1980 or later. It is interesting to examine trends in this data. For 

example, the ratios of the shallow to deep dose and neutron to deep dose for the 

respective recording periods are summarized in Table 5.3.4.2.1.2-2. These ratios exhibit 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 96 of 188

the expected response characteristics of the respective dosimeters based on laboratory 

studies.

Table 5.3.4.2.1.2-2 Hanford Ratio of Recorded Dose Components

Recording Period Ratio (Range)

Shallow/Deep Neutron/Deep

1950-56 1.6 (1.1 - 3.7) 0.003 (0 - 0.06)

1957-71 1.2 (1.1 - 1.7) 0.4 (0.1 – 0.7)

1972-95 1.3 (1.1 – 1.5) 0.6 (0.1 – 1.6)

The shallow dose response for the two-element film dosimeter used at Hanford prior to 

1957 is expected to be too high because of the significant over-response of this 

dosimeter to the low-energy photons prevalent in plutonium facilities (Figure 5.3.4.2-2). 

These data also show increasing levels of recorded neutron dose, relative to the deep 

dose, for each succeeding dosimeter design. 

While no formal report has been published establishing a neutron to photon ratio for 

Savannah River Workers, a review has been conducted as part of this Technical Basis 

Document to validate the assumption that the Hanford plutonium worker’s neutron to 

photon ratio was similar to that experienced by Savannah River workers. Neutron to 

photon ratios were evaluated for nine plutonium workers (5 from F-Area and 4 from HArea) totaling 38 monitoring years of data post 1971 (Table 5.3.4.2.1.1-3). While the 

average neutron to photon ratio was similar to the ratio observed by Fix et al (1996), 

there was a significant difference in the upper bound of the HB-Line. This variation 

could be the result of the higher specific activity of Pu-238 in the HB line compared to 

Pu-239 in the FB-line. 

Table 5.3.4.2.1.1-3 Post 1971 SRS Ratio of FB and HB Line Recorded Dose Components

Area Neutron/Deep Ratio Ratio (Average, Range)

(GM, GSD, 95th%) SRS Neutron/Deep Hanford Neutron/Deep

FB Line 0.36, 2.52, 1.65 0.52 (0.09 – 1.23) 0.6 (0.1 – 1.6)

HB Line 0.91. 2.84, 5.05 1.29 (0.05 – 3.10)

It is important to note that while an average was reported, the distribution of ratios more 

closely resembled a log normal distribution than a normal distribution. A distribution fit 

was conducted using the Kolmogrov –Smirnoff test statistic. The log normal distribution 

was demonstrated the best fit of the four distributions chosen (normal log normal, 

triangular, and uniform). For this reason, the geometric mean, geometric standard 

deviation and the corresponding 95th% is provided in Table 5.3.4.2.1.1-3. When 

incorporating uncertainty into the dose reconstruction, these values should be used in 

accordance with guidance on combining uncertainty provided in the External Dose 

Reconstruction Implementation Guideline (NIOSH 2002). 

Additional neutron to photon doses was found in the Monthly Works Technical Report 

(DuPont, 1965). A survey ratio of 0.66:1 was found for the HB-Line, and a ratio of 1:1 

was observed for the 235-F facility during this time frame. Brackenbush et al (1987) 

reported a neutron to gamma ratio of approximately 2:1. This variation in neutron to 

photon ratio is encompassed by the relatively large uncertainty distribution observed with 

worker data. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 97 of 188

5.3.4.2.2 Neutron Source Production Facilities (773-A)

A second area of neutron exposures at the Savannah River Site (SRS) resulted from the 

production of neutron sources such as Cf-252. Following irradiation in the reactors, Cf252 was separated. Unlike the plutonium facilities in which alpha particle interactions 

with light materials such as fluorine produce the neutrons (α,n reactions), Cf-252 

produces neutrons by spontaneous fission. As a result, the neutron energy spectra is 

considerably higher than the plutonium facilities. 

5.3.4.2.2.1 Neutron Energy Spectra

Measurements made by Brackenbush et al (1987) at the 773-A facility clearly 

demonstrated the shift in the neutron energy spectra. Using the same methodology 

described in the plutonium facility section, the neutron dose fractions were developed 

and are provided in Table 5.3.4.2.2.1-1. 

Table 5.3.4.2.2.1-1 Dose Fractions for Various Operations in 773-A Area

Neutron Energy 

Group Glovebox Shipping Cask Default

< 10 keV 0.018 0.015

10 – 100 keV 0.010 0.023

0.1 – 2 MeV 0.419 0.349

2 – 20 MeV 0.553 0.613

Claimant Favorable Dose Fractions

0.1 – 2 MeV 0.45 0.39 0.5

2 – 20 MeV 0.55 0.61 0.5

As with the plutonium facility dose fractions, a claimant favorable default value is 

provided when there is insufficient information to determine which dose fractions should 

be used. 

5.3.4.2.2.2 Neutron to Photon Ratio

As indicated with the plutonium workers, a review was conducted as part of this 

Technical Basis Document to evaluate the assumption that the Hanford plutonium 

worker’s neutron to photon ratio was similar to that experienced by Savannah River 

neutron source production workers. Neutron to photon ratios were evaluated for 6 

neutron source production workers from the 773-A area totaling 30 monitoring years of 

data post 1971 (Table 5.3.4.2.2.2-1). While the average neutron to photon ratio was 

slightly higher than the ratio observed by Fix et al (1996) for Hanford plutonium workers, 

there was a factor of 2 increase in the upper range. 

Table 5.3.4.2.2.2-1 Post 1971 SRS Ratio of 773-A Recorded Dose Components

Area Neutron/Deep Ratio Ratio (Average, Range)

(GM, GSD, 95th%) SRS Neutron/Deep Hanford Neutron/Deep

773-A 0.62, 2.29, 2.41 0.85 (0.10 – 3.83) 0.6 (0.1 – 1.6)
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 98 of 188

5.3.4.2.3 Reactor Facilities (100 Area)

The neutron dose to workers in SRS reactor facilities is very low in comparison to the 

measured photon and beta dose received from contaminated equipment and plant 

areas, spent reactor fuel, activated reactor components, and other areas containing 

fission and activation produces encountered during plant maintenance. However, there 

are a few areas around the reactors where a worker can receive neutron exposure, 

albeit a small fraction of the photon dose, during reactor operations. 

5.3.4.2.3.1 Neutron Energy

Brackenbush et al. (1987) recorded measurement data at the K-reactor door during 

operations is shown in Figure 5.3.4.2-3. The data indicated the lowest average neutron 

energy (170 keV) of all facilities measured at SRS. 

K-Reactor Behind Rector Shielding Wall

Neutron Energy 

10-7 10-6 10-5 10-4 10-3 10-2 10-1 100 101 102

Dose Equivalent (mrem)

10-5

10-4

10-3

10-2

10-1

100

101

0.07 mrem

0.26 mrem

1.95 mrem

0.0001 mrem

Figure 5.3.4.2-3 Neutron Energy Spectra Recorded Behind the K-Reactor Shield Door with 

Neutron Energy Groups Overlay.

Table 5.3.4.2.3.1-1 provides the dose fraction per neutron energy group. Unlike the 

previous two neutron exposure areas, the dose fraction from the intermediate energy 

neutrons was approximately 12% while the lower energy neutrons remained less than 

5%. Since the Radiation Effectiveness Factor (REF) for intermediate energy neutrons is 

greater than for lower energy neutrons, a claimant favorable simplification is to combine 

the dose fractions for lower and intermediate energy neutrons. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 99 of 188

Table 5.3.4.2.3.1-1 Dose Fractions for Heavy Water Reactors at SRS

Neutron Energy 

Group

Behind Shield 

Door Default

< 10 keV 0.030

10 – 100 keV 0.115

0.1 – 2 MeV 0.855

2 – 20 MeV 0.0001

Claimant Favorable Dose Fractions

10 – 100 keV 0.15 0.15

0.1 – 2 MeV 0.85 0.85

5.3.4.2.3.2 Neutron to Photon Ratio

Neutron exposures are expected to be very low compared to beta and photon exposures 

around the reactors because there is no neutron radiation when the reactors are 

shutdown for maintenance or refueling when most beta and photon dose to workers is 

received. Brackenbush et al (1987) reported both the neutron and photon dose rate at 

the F-Area and the K-Reactor. The neutron to photon dose rate ratio for the F-area was 

2:1, while the neutron to photon dose rate ratio for the K reactor was 1:1 or about half of 

the observed ratio from the F-Area. Using the simple dose-rate ratio reduction factor of 

2, the neutron to photon ratio for reactor area workers can be developed based on the 

personal dose ratio observed for the F-Area workers. Table 5.3.4.2.3.2-1 provides the 

F-Area neutron to photon ratio and calculated reactor area neutron to photon ratio. 

Table 5.3.4.2.3.2-1 Post 1971 SRS Ratio of Plutonium and Reactor Recorded Dose 

Components

Area

Neutron/photon Ratio Ratio (Average, Range)

(GM, GSD, 95th%) SRS 

Neutron/photon

Hanford 

Neutron/photon

Pu Facility 0.36, 2.52, 1.65 0.52 (0.09 – 1.23) 0.6 (0.1 – 1.6)

Reactor 0.18, 2.52, 0.82 0.26 (0.05 – 0.62) 0.26 (0.13 – 0.73)

Fix et al (1996) recalculated neutron doses and evaluated the neutron to photon ratio 

between the recalculated neutron dose and the photon dose using several methods. 

Based on this data, the most claimant favorable neutron to photon ratio for 100 area 

workers at Hanford between 1950 and 1961 ranged from 0.13 to 0.73 with a weighted 

average of 0.26. The SRS neutron to photon ratio is expected to be considerably lower 

than that observed from the Hanford reactors because there was minimal neutron 

radiation in routine work areas during reactor operation due to design (below grade) and 

shielding. Unlike the graphite reactors, in which workers could conduct work on the face 

of the reactor during operation, access to the lower plenum where the reactor core and 

cooling system were located was restricted. In addition, since most of the photon dose 

was received during refueling operations, increasing the denominator of the ratio, thus 

further lowering the ratio compared to plutonium facility workers. As a result, the neutron 

to photon ratio for SRS reactors is considered to be an upper bound of the actual 

neutron to photon ratio. 

Before applying the ratio listed in Table 5.3.4.2.3.1-1, careful consideration should be 

given to the energy employee’s work history. There are occupations around the reactor 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 100 of 188

areas in which no neutron dose would be expected (office or administrative staff). 

Generally only workers conducting work near the entrances to the lower plenum (Shield 

Door), the crane wash area, or the disassembly area have the potential for neutron 

exposure. There was also some neutron exposure in the assembly area due specific 

campaigns in which Pu-Al alloy targets were assembled. In the absence of claim – 

specific data to arrive at a neutron to photon ratio from Table 5.3.4.2.3.2-1 , a neutron to 

photon ratio of 0.104 should be applied to all non-administrative job categories. 

5.3.4.2.4 Other Facilities

5.3.4.2.4.1 Calibration Facilities (736-A Area)

The neutron exposure spectra for workers at the SRS calibration Facility cover the entire 

energy spectra. Although Brackenbush et al (1987) measured multiple spectra at this 

facility, only two that represent the extremes are discussed in this analysis. The first is 

the routine calibration of dosimeters using a PuF4 source in free air. The second is the 

spectra recorded using a PuBe source at a tank containing 150 mm of D2O. Table 

5.3.4.2.3.1-1 demonstrates the wide variation in the resultant dose fractions. Since the 

relative use of these different sources is not known and the Radiation Effectiveness 

Factor (REF) is the largest (claimant favorable) for fission spectrum neutrons, the default 

was chosen based on the PuF4 source. 

The neutron to photon ratio for the calibration facility is not currently known. However, 

since intermittent exposure is considered the most probable scenario, the neutron to 

photon ratio is not expected to be greater than that observed at the 773-A facility. A 

reasonable and claimant favorable assumption would be to use the neutron to photon 

ratio for the 773-A facility. 

Table 5.3.4.2.3.1-1 Dose Fractions at the Calibration Facility (736-A)

Neutron Energy 

Group

PuF4 Source w/o 

Drum 

PuBe Source at 

150 mm D2O Default

< 10 keV 0.008 0.127

10 – 100 keV 0.007 0.046

0.1 – 2 MeV 0.819 0.374

2 – 20 MeV 0.166 0.453

Claimant Favorable Dose Fractions

0.1 – 2 MeV 0.83

2 – 20 MeV 0.17

5.3.4.2.4.2 Fuel Fabrication Area (321-M) 

For certain campaigns, Pu-Al alloy targets were fabricated in the 321-M area. Figure 

5.3.4.2-4 reproduced from Brackenbush et al (1987), compares the neutron spectra of Pu-Al 

to various other neutron sources. As can be seen from the graph, Pu-Al alloy has a similar 

neutron energy spectra to PuF4. Since the Pu-Al alloy is similar to a bare PuF4 source, the 

 

4 Claimant favorable factor determined during June 19, 2003 teleconference among NIOSH, NIOSH contractor and 

SRS technical staff for default use based on minimal neutron dose in SRS reactors, absolutely, and certainly in 

comparison with the associated photon dose. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 101 of 188

dose fraction for the PuF4 source w/o Drum should be used (Table 5.3.4.2.3.1-1). Neutron 

exposures in the Fuel Fabrication Area are expected to be intermittent and were not 

continuous over the operation of the SRS, thus a neutron to photon ratio should not be 

used. Based on dosimetry records from SRS, it appears that during the Pu-Al campaigns, 

workers were monitored for neutron exposure. Since the Pu-Al alloy spectra is of similar 

energy to PuF4, approximately 90% of the spectra is greater than the 500 keV threshold for 

NTA film. As a result, the NTA film measurements are expected to be reasonably accurate 

to within parameters discussed in the External Dose Reconstruction Implementation 

Guideline (NIOSH 2002). 

Figure 5.3.4.2-4 Comparison of Various Neutron Spectra from different Plutonium 

Compounds and Alloys (Brackenbush et al, 1987)

5.3.5 SRS Workplace Neutron Dose Uncertainty

As noted previously, in 1987, SRS contracted the Pacific Northwest National Laboratory 

(PNNL) to conduct neutron spectra and dose measurements at selected SRS facilities 

(Brackenbush et al 1987) similar to measurements made by PNNL at Hanford 

(Brackenbush et al 1991, Endres et al 1996, Scherpelz et al 2000, Fix et al 1981, 1982). 

These neutron measurements were made using several methods to measure neutron 

dose including Tissue Equivalent Proportional Counters (TEPCs), portable neutron 

survey instruments, and the SRS Hoy TLND. In addition, track etch dosimeters from 

Hanford and Lawrence Livermore National Laboratory (LLNL) were used since they 

were being studied for potential use in combination personnel neutron dosimeter 

systems. Energy spectrum measurements were made using multisphere (Bonner) 

sphere spectrometers, 3

HE spectrometers, and NE-213 liquid scintillation spectrometers. 

As reported in the PNNL report, measurements with the TEPC, multisphere system and 

3

HE spectrometer were in general agreement. The TLND agreed within about 30% for 

most measurement locations along the plutonium production lines and storage areas. 

The TLND was within a factor of 3 (i.e., 0.3 to 3) for the extremes in neutron energy 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 102 of 188

spectra encountered at the K-reactor door (i.e., highly thermalized field) and for a 

californium shipping cast (i.e., where most lower energy neutrons had been removed). 

Over long time periods, workers would generally be expected to be involved in several 

different exposure profiles that will serve to minimize the extremes identified. These 

results are indicative of the technical difficulties to accurately measure neutron dose in 

the workplace. Table 5.3.4.2-3 presents a summary of common workplace neutron 

dosimeter performance characteristics. 

Table 5.3.4.2-3 Common Workplace Neutron Dosimeter Hp(10) Performance(a)

Parameter Description Workplace Bias(a)

Workplace 

neutron energy 

spectra

SRS NTA dosimeter response 

decreases and SRS TLD 

response increases with 

decreasing neutron energy.

Depends upon workplace neutron 

spectra. NTA recorded dose of record 

likely too low. Article by Hoy presented 

in Vallario et al (1969) identified ratios of 2 

to 4 under-response. 

Exposure 

Geometry NTA Dosimeter response 

increases with increasing 

exposure angle and TLD 

response decreases with 

increasing exposure angle. 

NTA recorded dose likely too high since 

dosimeter response is higher angles other 

than A-P. TLD recorded dose of record 

likely too low since dosimeter response is 

lower at angles other than A-P. Effect is 

highly dependent upon neutron energy. 

Dosimeter 

Placement

SRS apparently instructed 

workers to wear NTA dosimeter 

facing the workers body to 

enhance dosimeter response 

(i.e., to respond to backscatter 

neutron radiation).

Recorded dose of record likely too low. 

Reflected neutrons are expected to be 

less-than the NTA energy threshold 

identified as 0.5 MeV by SRS. 

Mixed low 

energy photon 

and neutron 

radiation

SRS NTA dosimeter responds 

significantly to lower-energy 

photons that can result in fogging 

of the NTA film.

NTA recorded dose of record likely too 

low. SRS NTA dosimeters selected to 

minimize photon response. SRS TLDs 

have capability to remove photon dose 

contribution. 

a. Judgment based on SRS dosimeter response characteristics and workplace radiation fields.

b. Recorded dose compared to Hp(10).

Taylor et al (1995) reported that SRS participated in Personal Dosimetry 

Intercomparison Studies starting in 1974. SRS dosimetry demonstrated an average 

error of approximately 40% in the first intercomparison. According to Taylor et al (1995), 

a marked improvement was noted and the average error was 13.9% in 1985. The SRS 

site has noted a trend of improving neutron dosimetry through the years. Measurements 

of TLND performance at SRS in 1987 (Brackenbush et al., 1987) indicate that the SRS 

measured neutron dose with the TLND (beginning 1 January 1971) is reasonably 

correct. For dose reconstruction under EEOICPA, a claimant favorable standard error 

estimate of 50% should be made for neutron dosimetry between 1971 and 1985. 

Starting in 1985, a standard error of 15% should be assumed in the neutron recorded 

dose. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 103 of 188

5.4 Adjustments to Recorded Dose

Adjustments to the SRS reported dose are necessary to arrive at a claimant-favorable 

dose considering the uncertainty associated primarily with the complex SRS workplace 

radiation fields and exposure orientations. 

5.4.1 Photon Dose Adjustments

Taylor et al (1995) identified calibration corrections to the reported SRS TLD-Deep 

photon dose for the periods prior to 1986 and 1987, respectively, to estimate the Hp(10)

dose. Photon dose corrections summarized in Table 5.4.1-1 are necessary to calculate 

an adjusted dose that is claimant favorable considering historical technological 

limitations of SRS dosimeters. For the years 1987 to the present, no correction is 

needed, the reported dose is Hp(10) equivalent.

Table 5.4.1-1 Adjustments to Reported SRS Deep Photon Dose

Time 

Period Dosimeter Facility Step Adjustment to Reported Dose

Prior to 

1/1/1986

All 

beta/photon 

dosimeters

All facilities A Multiply reported TLD-Deep photon 

dose by a factor of 1.119 to estimate 

Hp(10).

For the year 

1986

TLD 

beta/photon 

dosimeter

All facilities B Multiply reported TLD-Deep photon 

dose by a factor of 1.039 to estimate 

Hp(10).

5.4.2 Neutron Dose Adjustments

The Savannah River Site (SRS) incorporated the energy variation of the dose equivalent 

into their calibration methodology. As a result, the recorded dose equivalent (DER) is a 

combination of all neutron energies. In order to calculate the probability of causation, 

the recorded neutron dose must be separated into neutron energy groups as discussed 

in section 5.3.4.2 and subsequently converted to ICRP 60 (1990) methodology. 

5.4.2.1 Neutron Weighting Factor 

Adjustment to the neutron dose is necessary to account for the change in neutron quality 

factors between historical and current scientific guidance as described in NIOSH (2002). 

At SRS, a single quality factor was not used, but rather a continuum of quality factors 

based on neutron energy that varies up to 11 at 1 MeV (Singh 2002). The quality factor 

was incorporated into the calibration methodology which used flux to dose-rate 

conversion factors for varying neutron energies (Brackenbush 1987, Taylor 1995). 

According to site personnel (Crase 2003), the SRS flux to dose-rate conversion factors 

were based on NCRP 38 (1971). NCRP 38 (1971) lists both flux to dose-rate conversion 

factors and associated quality factors that vary from 2 at energies less than 1 keV to 11 

at 1 MeV. In order to convert from NCRP 38 quality factors to ICRP 60 (1990) radiation 

weighting factors, a curve was fit describing the neutron quality factors as a function of 

neutron energy. The average quality factor for each neutron energy group was 

developed by integrating the area under curve and dividing by the neutron energy range 

as shown in equation 5.2. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 104 of 188

)1.00.2(

)(

( )

0.2

1.0

0.21.0, − =

∫ − Range

dEEQ

EQ

f

n MeV (5.2)

Table 5.4.2-1 provides a summary of historical changes in the quality factors and the 

average NCRP 38 quality factor for the neutron energy groups used in dose 

reconstruction. 

Table 5.4.2-1 Historical Neutron Quality or Weighting Factors 

Neutron 

Energy

(MeV)

Historical 

Dosimetry 

Guideline(a)

NCRP 38

Quality 

Factors(b)

Average 

Quality Factor 

used at SRS

ICRP 60 Neutron 

Weighting factor, 

wr 

(c)

2.5 x 10-8 3 2 

2.35 5 

1 x 10-7

10

2 

1 x 10-6 2 

1 x 10-5 2 

1 x 10-4 2 

1 x 10-3 2

1 x 10-2 2.5 5.38 10

1 x 10-1 7.5

5 x 10 10.49 20 -1 11

1 11

2 10

7.56 10

2.5 9 

5 8 

7 7 

10 6.5

14 7.5

20 8

40 7 Not applicable 5 

60 5.5

a. Trilateral meeting in 1949 radiation protection guidelines (Fix et al 1994). 

b. Recommendations of NCRP 38.

c. ICRP 60 (1990) 

5.4.2.2 Neutron Correction Factor

Table 5.4.2-1 provides the average quality factor for the four neutron energy groups 

which encompass SRS neutron exposures. The neutron dose equivalent correction 

factor can be calculated by dividing the dose fractions from section 5.3.4.2 for each 

neutron energy group (Df(En)) by the corresponding energy specific average NCRP 38 

quality factor (Q(En)) and then multiplying by the ICRP 60 radiation weighting factor (wR) 

as shown in equation 5.3. 

 

R

n

nf

nf w

EQ

ED

EC ×=

)(

)( )( ( 5.3)
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 105 of 188

Table 5.4.2.2-1 summarizes the default neutron dose distributions by energy for each 

neutron exposure area. By multiplying the recorded neutron dose by the area specific 

correction factors the neutron dose equivalent is calculated. For example consider a 

1000 mrem recorded neutron dose by an energy employee working at the HB-Line, the

corrected neutron dose is 1144 mrem from neutrons between 0.1-2.0 MeV and 529 

mrem from neutrons with energy between 2 and 20 MeV. Thus the corrected neutron 

dose is a total of 1673 mrem. These corrections should be applied to measured dose, 

missed dose and dose determined based on a neutron to photon ratio. Table 5.4.2.2-1 

summarizes the default dose fractions by energy and the associated ICRP 60 correction 

factors for each neutron exposure area. 

Table 5.4.2.2-1 Summary of Dose Fractions and associated ICRP 60 correction factors

Process Description / Buildings

Operations Neutron 

Energy

Default 

Dose 

Fraction 

(%)

ICRP 60 

Correction 

Begin End Factor

Fuel 

Fabrication

Produced reactor fuel and target assemblies. Intermittent production mixed oxide fuel and plutonium – 

aluminum (Pu-Al) alloy targets.

(321-M)

0.1-2 MeV

2-20 MeV

83%

17%

1.58

0.23

Reactors

During Reactor Operation: Low level neutron exposure through shielding. In addition, some intermittent 

neutron exposure from mixed oxide fuel and targets. 

(105C) 3/1955 1/1987

10-100 keV

0.1 – 2 MeV

15%

85%

0.28

1.62

(105K)

10/1954 1/1988

1992 7/1992

(105L)

7/1954 2/1968

1985 6/1988

(105P) 2/1954 8/1988

(105R) 12/1953 6/1964

Plutonium 

Production

Plutonium Finishing Process: Plutonium enters the process as PuF4 and is then fired into production 

pucks. Work is primarily conducted in glove boxes with predominant close anterior exposure to workers. 

Radiation levels at the beginning of the process are fairly constant while levels at end of process are 

closely related to production levels. 

Plutonium Finishing Process

(221-H, B-Line)

11/1954 1988

0.1-2 MeV

2-20 MeV

60%

40%

1.14

0.53

Plutonium Finishing Process

(221-F, B-Line)

10/1955 1988 0.1-2 MeV 100% 1.91

Plutonium Storage

0.1-2 MeV

2-20 MeV

86%

14%

1.64

0.19

Production Control Laboratory

(772-F)

1955 0.1-2 MeV 100% 1.91

Plutonium Fuel Facility

(235-F)

1955 0.1-2 MeV 100% 1.91

Radionuclid

e 

Production 

and 

Calibration

Radiation characteristics are highly dependent upon the radionuclide being produced. SRS has produced 

several radionuclides including 238Pu, 239Pu, americium, curium, californium, etc. At the calibration facility 

radiological instruments are calibrated using a variety of sources with different neutron energy spectra. 

 

Californium Production Facility 

(773-A)

0.1-2 MeV

2-20 MeV

50%

50%

0.95

0.66

Calibration Facility (736-A) 0.1-2 MeV

2-20 MeV

83%

17%

1.58

0.23
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 106 of 188

5.5 Missed Dose

There is undoubtedly missed recorded dose for SRS workers. Analysis of the missed 

dose has been separated according to photon and neutron missed dose. 

5.5.1 Photon Missed Dose

Missed photon dose for SRS workers would occur where: 1) there is no recorded dose 

because workers were not monitored or the dose is unavailable and 2) a zero dose is 

recorded for the respective SRS dosimeter systems for any dosimeter response lessthan the minimum detection level (MDL). Methods to be considered when there is no 

recorded dose for a period within a working career have been examined by Watson et al 

(1994). Generally, estimates of the missed dose can be made using dose results for coworkers or using the recorded dose prior to and after the period of missed dose. 

However, these situations do require careful examination. The missed dose for 

dosimeter results less-than the MDL is particularly important for earlier years when 

MDLs were higher and dosimeter exchange was more frequent. NIOSH (2002) 

describes options to calculate the missed dose. One option is estimate a claimant 

favorable maximum potential missed dose where the Minimum Detection Level (MDL) is 

multiplied by the number of zero dose results providing an estimate of the maximum 

potential missed dose. The following sections describe the potential missed photon 

dose corrections according to facility/location, dosimeter type, year and energy range.

5.5.1.1 Facility/Location

The potential missed photon dose for the laboratory determined MDL and exchange 

frequency using NIOSH (2002) is presented in Table 5.5.1-1. 

Table 5.5.1-1 Missed Photon Dose According to SRS Facility.

SRS Facility 

Type Dosimeter Period of Use

Exchange 

Frequency

MDL 

(mSv)(a)

Max. Annual 

Missed Dose

(mSv)(b)

Fuel Fabrication, 

SRS Reactors, 

Processing 

Plants, Waste 

Handling

ORNL Two-Element 

film Dosimeter

Prior to 3/52 Weekly 0.4 20.8 

SRS Two Element 

film Dosimeter

After 3/52 through 

9/30/1957

Weekly 0.4 20.8 

10/1/1957 through 

11/8/1959

Biweekly 0.4 10.4 

Above plus 

plutonium and 

nuclide 

production 

SRS Multi-Element 

Film Dosimeter

11/9/1959 through 

12/31/1964

Biweekly 0.4 10.4

1/1/1965 through 

12/31/1965

4-week 0.4 5.2

1/1/1966 through 

3/31/1970

Monthly 0.4 4.8

All SRS SRS TLD 4/1/1970 through 

6/30/1983

Monthly 0.15 1.8

Panasonic TLD 7/1/1983 to 2003 

(ongoing)

Monthly 0.05 0.6

Panasonic TLND 1/1/1995 to 2003 

(ongoing)

Monthly 

(n=12)

0.15 0.9

a. Estimated film dosimeter detection levels based on NIOSH (1993), NRC (1989), and Wilson et al (1990). 

TLD detection levels from Taylor et al (1995) and SRS (1993). 

b. Maximum annual missed dose calculated from OCAS-IG-001 (NIOSH 2000). 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 107 of 188

5.5.1.2 Dosimeter Type

The MDLs for the respect SRS beta and photon dosimeters normally cited are based on 

laboratory irradiations. The actual MDLs are greater than these values because of 

additional uncertainty in actual field use. Reasonable MDLs are shown in Table 5.5.1-2 

for most applications based on NIOSH (1993), NRC (1989), and Wilson et al (1990). 

The TLD detection levels from Taylor et al (1995) and SRS (1993). 

Table 5.5.1-2 Missed Photon Dose According to Dosimeter Type

Dosimeter

Type

Period of Use MDL

(mSv)

Exchange 

Frequency

Max. Annual 

Missed Dose

(mSv)

Two 

Element 

Film

Prior to 3/52 through 9/30/1957 0.4 Weekly (n = 52) 20.8

10/1/1957 through 11/8/1959 0.4 Biweekly (n=26) 10.4

MultiElement 

Film

11/9/1959 through 12/31/1964 0.4 Biweekly (n=26) 10.4

1/1/1965 through 12/31/1965 0.4 4-week (n=13) 5.2

1/1/1966 through 3/31/1970 0.4 Monthly (n=12) 4.8

TLD

4/1/1970 through 6/30/1983 0.15 Monthly (n=12) 1.8

7/1/1983 to 2003 (ongoing) 0.05 Monthly (n=12) 0.6

0.05 Quarterly (n = 4) 0.2

5.5.1.3 Year

Analysis of the missed photon dose according to year (actually by period according to 

dosimeter type and exchange) may be needed to evaluate some claim information 

particularly if only annual dose data are available. Table 5.5.1-3 summarizes the 

potential missed dose. 

Table 5.5.1-3 Missed Photon Dose Adjustments to Recorded Deep Dose

Period of Use MDL 

(mSv)

Exchange 

Frequency

Max. Annual 

Missed Dose

(mSv)

Prior to 3/52 through 9/30/1957 0.4 Weekly (n = 52) 20.8

10/1/1957 through 11/8/1959 0.4 Biweekly (n=26) 10.4

11/9/1959 through 12/31/1964 0.4 Biweekly (n=26) 10.4

1/1/1965 through 12/31/1965 0.4 4-week (n=13) 5.2

1/1/1966 through 3/31/1970 0.4 Monthly (n=12) 4.8

4/1/1970 through 6/30/1983 0.15 Monthly (n=12) 1.8

7/1/1983 to 2003 (ongoing) 0.05 Monthly (n=12) 0.6

5.5.1.4 Energy Range

An estimate of the missed photon dose by energy range is possible based on the type of 

facility and predominant radionuclides such as intermediate (>100 keV) energies for all 

facilities handling activation and fission product nuclides, primarily lower (<100 keV) 

energy photons for plutonium facilities and lower energy photons for uranium fuel 

fabrication facilities. The recorded dose from the dosimeter response does not typically 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 108 of 188

provide sufficient information to estimate discrete energy ranges. It is possible to 

examine the energy response characteristics of the respective multi-element dosimeters 

but this analysis does not recognize the substantial uncertainties present in the 

workplace associated with shielding, radiation scattering and mixed radiation fields.

5.5.2 Neutron Missed Dose

There is potential for significant missed dose among workers at the plutonium finishing 

facilities and the neutron source production area. There is also some potential for 

missed dose from work around the reactors and at the fuel fabrication facility. The 

approach used to calculate the neutron missed dose can be divided into two time 

periods. The first is before 1971 when neutron track emulsion type A (NTA) film was 

used and the second is when thermoluminescent dosimeters were used. Table 5.5.2-1 

provides a summary of the reported limits of detection by the Savannah River Site 

(SRS).

Table 5.5.2-1 Summary of SRS Neutron Dosimeter Limits of Detection (LOD)

Dosimeter Type

Dates of Use Exchange 

Frequency

Reported 

LOD

(mrem)

Estimated 

LOD

(mrem)

Max. Annual 

Missed Dose

Begin End (mrem)

ORNL NTA 1951 8/3/1953 Weekly

(n = 52) ~50(a) ~50 2600

SRS NTA 8/3/1953 7/13/1960 Weekly

(n = 52) 30 ~40(b) 2080

SRS NTA 7/14/1960 12/31/1970 Biweekly 

(n=26) 30 ~40(b) 1040

SRS Hoy TLND 1/1/1971 12/31/1994 Monthly 

(n=12) 10(c) 20 240

Panasonic TLND 1/1/1995 Present Monthly 

(n=12) 15 15 180

(a) Based on Wilson et al (1990)

(b) Correction for estimated 25% under response of NTA film for some facilities

(c) Minimum Recorded Dose not necessarily LOD

5.5.2.1 Prior to 1971

Due to the degraded energy spectra at most SRS facilities, the NTA film generally under 

responded due to the energy threshold of approximately 500 keV. The magnitude of the 

NTA film under response ranges from about 10% at the 773-A facility to over 90% 

around the reactors. According to Taylor et al. (1995), the SRS calibration method for 

NTA film might have accounted for some of this under response, however, until this can 

be further evaluated, the claimant favorable approach is to assume that it was not 

incorporated into the calibration methodology. As a result of the inability of NTA film to 

measure neutrons below 500 keV, the “true” limit of detection varies by energy spectra 

or facility. In the case of reactor operations, the limit of detection is virtually 

meaningless, while for the Californium Production facility (773-A), the reported limit of 

detection is probably reasonable however might be under reported by 10%. 

Due to the uncertainty in whether an energy employee’s NTA badge would respond to 

the workplace neutron spectra, using a ratio to the measured photon dose is 

recommended as a claimant-favorable option to reconstruct an individual worker neutron 

dose. This is based on the fact that for routine neutron exposure, the neutron exposure 

is essentially always accompanied with measurable photon exposure. Table 5.5.2-2 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 109 of 188

provides a summary of the recommended neutron to photon ratios for the different 

process areas. As can be determined from the table, the recommended method to apply 

the ratio is as a log normal distribution using the geometric mean and geometric 

standard deviation. Caution, however, should be employed when using Table 5.5.2-2 

and consideration should be given to an individual energy employee’s neutron to photon 

ratio if there is sufficient individual data. For example if an individual has recorded 

neutron dose either prior to or after the estimation period that has a higher neutron to 

photon ratio than what is provided in the table, then the higher neutron to photon ratio 

should be used provided that the work assignments were similar for the two time 

periods. The values in Table 5.5.2-2 are intended to provide guidance when there is 

known technical limitations in the neutron dosimetry technology or no neutron monitoring 

data. The data were developed by evaluating multiple worker years, thus for a given 

time period, the “true” ratio might be more or less than the geometric mean. Due to 

potential for photon missed dose effecting the neutron to photon ratio, when applying the 

ratio methodology above, the photon dose should be the combination of both the 

measured photon dose (Dosimeter Dose) and the photon missed dose. 

When overestimating dose reconstructions, the upper 95% ratio can be used as a 

claimant-favorable worst case ratio and the associated neutron uncertainty can be 

omitted. In underestimating dose reconstructions, the geometric mean can be used as a 

best case ratio. 

Table 5.5.2-2 Summary of neutron to photon ratios by process

Process Description / 

Buildings

Operations Neutron to Photon Ratio

Begin End

Geometric 

Mean

(GM)

Geometric 

Standard 

Deviation (GSD)

Upper 

95th% 

Fuel 

Fabrication

Produced reactor fuel and target assemblies. Intermittent production mixed oxide fuel and plutonium – 

aluminum alloy targets.

(321-M) (a) (a) (a)

Reactors During Reactor Operation: Low level neutron exposure through shielding. In addition, some intermittent 

neutron exposure from mixed oxide fuel and targets. 

(105C) 3/1955 1/1987

0.18 2.52 0.82

(105K)

10/1954 1/1988

1992 7/1992

(105L)

7/1954 2/1968

1985 6/1988

(105P) 2/1954 8/1988

(105R) 12/1953 6/1964

Plutonium 

Production

Plutonium Finishing Process: Plutonium enters the process as PuF4 and is then fired into production 

pucks. Work is primarily conducted in glove boxes with predominant close anterior exposure to workers. 

Radiation levels at the beginning of the process are fairly constant while levels at end of process are 

closely related to production levels. 

Plutonium Finishing Process

(221-H, B-Line)

11/1954 1988 0.91 2.84 5.05

Plutonium Finishing Process

(221-F, B-Line)

10/1955 1988 0.36 2.52 1.65

Plutonium Storage

Production Control 

Laboratory

(772-F)

1955
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 110 of 188

Process Description / 

Buildings

Operations Neutron to Photon Ratio

Begin End

Geometric 

Mean

(GM)

Geometric 

Standard 

Deviation (GSD)

Upper 

95th% 

Plutonium Fuel Facility

(235-F)

1955

Radionuclide

Production 

and 

Calibration

Radiation characteristics are highly dependent upon the radionuclide being produced. SRS has 

produced several radionuclides including 238Pu, 239Pu, americium, curium, californium, etc. At the 

calibration facility radiological instruments are calibrated using a variety of sources with different neutron 

energy spectra. 

 

Californium Production 

Facility (773-A)

0.62 2.29 2.41

Calibration Facility (736-A) (b) (b) (b)

(a) During Pu-Al target production, workers with potential for neutron exposure wore NTA film, thus LOD values listed in 

table 5.5.2-1 should be used. 

(b) Neutron to photon ratio for the calibration facility is not expected to be greater than the ratio observed in the 

californium production facility. As a claimant favorable approximation the Californium Production facility ratio can be 

used. 

5.5.2.2 Post 1971

Since the introduction of thermoluminescent neutron dosimeters at the site, neutron 

exposures have been fairly well monitored and the recorded data is reasonably 

accurate. As a result the guidance on missed dose provided on the External dose 

Reconstruction Implementation Guideline (NIOSH 2002) should be followed in 

accordance with the LOD values presented in Table 5.5.2-1. 

5.6 Organ Dose 

Once the Hp(10) adjusted doses have been calculated for each year, these values are 

used according to NIOSH (2002) to calculate the organ dose distribution. 

5.6.1 Organ Dose Conversion Factors

NIOSH (2002) describes the methodology used to calculate the organ dose distribution 

for the respective radiation types using identified exposure geometries. The selection of 

the worker orientation is important to the calculation of the organ dose. Examples of 

common exposure orientations are provided in the NIOSH Implementation Guide (2002), 

Table 4.2. Unfortunately, there is no definitive process to determine the exposure 

geometry for each worker. Table 5.6.1-1 lists proposed default options based on 

judgment of claimant favorable exposure geometries for long-term SRS workers. 

Table 5.6.1-1 Default Exposure Geometries to calculate Organ Dose 

Claim Status Job Category (b) Exposure Geometry Percentage (c)

Likely non-compensable All AP 100%

Compensable - Workers All AP 50%

ROT 50%

Compensable - Supervisors All 

AP 50%

ISO 50%

Attachment E presents a discussion of the use of the various parameters presented in 

this section that will aid the dose reconstructionist in preparing dose reconstructions.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 111 of 188

6.0 REFERENCES

Arnett, M. W., 1993, Savannah River Site Environmental Data for 1992, WSRC-TR-93-

077, Westinghouse Savannah River Company, Aiken, SC.

Biles, M. B., 1972 – Letter to T. A. Nemzak (Ad Hoc Technical Committee Finds), Pacific 

Northwest National Laboratory, Richland, WA, 23 November.

Bebbington, W.P., (1990) History of Du Pont at the Savannah River Plant, E.I. du Pont 

de Nemours and Company, Wilmington, Delaware.

Brackenbush, L. W., W. V. Baumgartner and J. J. Fix. 1991. Response of TLD Albedo 

and Nuclear Track Dosimeter Exposed to Plutonium Sources. PNL-7881, Pacific 

Northwest Laboratory, Richland, Washington.

Brackenbush, L.W., G.W.R. Endres, J. M. Selby, and E.J. Vallario. 1980. Personnel 

Neutron Dosimetry at Department of Energy Facilities. PNL-3213. Pacific Northwest 

Laboratory, Richland, WA 99352.

Brackenbush, L.W., K.L. Soldat, D. L. Haggard, L. G. Faust and P. L. Tomeraasen. 

1987. Neutron Dose and Energy Spectra Measurements at Savannah River Plant. 

PNL-6301. Pacific Northwest Laboratory, Richland, WA 99352.

Brown, K.T., 2001 – Letter to Mr. Timothy D. Taulbee, National Institute for Occupational 

Safety and Health: Occupational X-Rays at the Savanna River Site.

Buschbom, R.L. and E.S. Gilbert. 1993. Summary of Recorded External Radiation 

Doses for Hanford Workers 1944-1989. PNL-8909, Pacific Northwest Laboratory, 

Richland, WA.

CDC 2001 – Savannah River Site Dose reconstruction Project Phase II: Source Term 

Calculation and Ingestion Pathway Data Retrieval, Evaluation of Materials Released 

from the Savannah River Site, RAC Report No. 1-CDC-SRS-1999, Centers for Disease 

Control and Prevention, Atlanta, GA, April. 

Cardarelli, J.J., II, 2000. A Potential Consequence of Excluding Work-Required A-Ray 

Exposures When Computing Cumulative Occupational Radiation Dose at a Uranium 

Enrichment Plant. A dissertation submitted to the Division of Research and Advance 

Sciences of the University of Cincinnati in partial fulfillment of the requirements for the 

degree of Doctorate of Philosophy, Cincinnati, OH.

Cardarelli, J. J., Spitz, H. Rice, C, Buncher, R Elson, H., and Succop, P. (2002) 

Significance of Radiation Exposure From Work-Related Chest X-Rays for 

Epidemiological Studies of Radiation Workers AMERICAN JOURNAL OF INDUSTRIAL 

MEDICINE 42:490–501
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 112 of 188

Crase, K. W. (2003), Personal Communication with Tim Taulbee, June 2003

Cummins, C.L., C.S. Hetrick, and D.K. Martin. 1991. Radioactive Releases from the 

Savannah River Plant 1954–1989, Environmental Protection Department Summary.

WSRC-RP-91-684. 

Currie, L. A. 1968. "Limits for Qualitative Detection and Quantitative Determination," 

Analytical Chemistry, 40(3), 586-593, March 1968

DOE/EM-3019, January 1997, Linking Legacies, U.S. Department of Energy

Department of Energy (DOE). 1986. Department of Energy Standard for the 

Performance Testing of Personnel Dosimetry Systems. DOE/EH-0027. US Department 

of Energy, Washington, D.C.

Department of Energy (DOE). 1996. Plutonium: The First 50 Years United States 

Plutonium Production, Acquisition, and Utilization from 1944-1994, DOE/DP-0137, U. S. 

Department of Energy, Washington, D.

Du Pont, 1954, 1957, 1961, 1961, Administrative Procedures Manuals

Dupont, (1965) Progress Report: Works Technical Department - August 1965, DPSP 65-

1-8, E.I. du Pont de Nemours & Company, Savannah River Plant.

Dupont, (1961). Operating Procedures: Health Physics Personnel Meters, DPSOP-45, 

Rev. 2, E.I. du Pont de Nemours & Co., Inc, Savannah River Plant, Aiken, South 

Carolina.

Dupont, (1961). Operating Procedures: Health Physics Personnel Meters, DPSOP-45, 

Rev. 3, E.I. du Pont de Nemours & Co., Inc, Savannah River Plant, Aiken, South 

Carolina.

Du Pont, (1954). Operating Procedures: Health Physics Personnel Meters, DPSOP-45, 

Rev. 0, E.I. du Pont de Nemours & Co., Inc, Savannah River Plant, Aiken, South 

Carolina.

Du Pont, (1957). Operating Procedures: Health Physics Personnel Meters, DPSOP-45, 

Rev. 1, E.I. du Pont de Nemours & Co., Inc, Savannah River Plant, Aiken, South 

Carolina. 

Endres, A. W., L. W. Brackenbush, W. V. Baumgartner, J. J. Fix and B. A. Rathbone. 

(1996). Response of the Hanford Combination Neutron Dosimeter in Plutonium 

Environments. PNL-10561, Pacific Northwest Laboratory, Richland, Washington.

EPA (1993), External Exposure to Radionuclides in Air, Water, and Soil, Federal 

Guidance Report No. 12, Office of Radiation and Indoor Air, U.S Environmental 

Protection Agency, Washington, DC.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 113 of 188

Erickson, Marilyn, January 22, 2003; Telephone Conversation. Historical Chest X-Ray 

Equipment and Methods at the Savannah River Site

Fix, JJ, L. Salmon, G. Cowper and E. Cardis. A Retrospective Evaluation of the 

Dosimetry Employed in an International Combined Epidemiologic Study. Radiation. 

Protection. Dosimetry. 74, 39-53, 1997.

Fix, J. J, E.S. Gilbert and W. V. Baumgartner. 1994. An Assessment of Bias and 

Uncertainty in Recorded Dose from External Sources of Radiation for Workers at the 

Hanford Site. PNL-10066, Pacific Northwest Laboratory, Richland, WA.

Fix, J. J., Wilson, R. H. and Baumgartner, W. V., (1996) Battelle. Retrospective 

Assessment of Personnel Neutron Dosimetry for Workers at the Hanford Site, PNNL11196, Pacific Northwest National Laboratory, Richland, Washington.

Fix, et al, 1997, The International Agency for Research on Cancer (IARC) Three County 

Combined Study

Garmin 2001, MapSource Version 4.03, Garmin Corporation, Olathe, KS

Gilbert, E.S., and J.J. Fix. 1996. Laboratory Measurement Error in External Dose 

Estimates and Its Effects on Dose-Response Analyses of Hanford Worker Mortality Data. 

PNNL-11289, Pacific Northwest National Laboratory, Richland, WA.

Hamby, D.M., and Parker, M.J., 1991, Gaussian Dispersion and Dosimetric Modeling 

Sensitivity to Area Specific 1982-86 Meteorological Data Collected at the Savannah 

River Site, WSRC-RP-91-9009, Westinghouse Savannah River Company, Savannah 

River Site, Aiken, SC

Harvey, R. P., D. M. Hamby, 2001, The Creation of an Historical Database for Dose 

Reconstruction, WSRC-TR-2001-00275, Westinghouse Savannah River Company, 

Savannah River Site, Aiken, SC

Hoy, J. E. 1972. Personnel Albedo neutron Dosimeter with Thermoluminescent 7Li and 

6Li. DP-1277. E.I. du Pont de Nemours and Company, Inc., Savannah River Plant, 

Aiken, South Carolina. 

Hoy, J. E. 1980. Thermoluminescent Dosimeters for Personnel Neutron 

Monitoring. DPST-70-533. E.I. du Pont de Nemours and Company, Inc., 

Savannah River Plant, Aiken, South Carolina.

 La Bone, Thomas R., October 1, 1996, Facility Descriptions (U), ESH-HPT-960197

ICRP, (1975), Reference Man: Anatomical, Physiological and Metabolic Characteristics, 

ICRP Publication 23, Pergamon Press, Oxford, England. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 114 of 188

ICRP, (1983) Protection of the Patient in Diagnostic Radiology, ICRP Publication 34, 

Pergamon Press, Oxford, England. 

ICRP, (1990) Recommendations of the International Commission on Radiological 

Protection .ICRP Publication 60, Annals of the ICRP, Vol.21, Pergamon Press, Oxford 

England

ICRP (1994), Dose Coefficients for Intakes of Radionuclides by Workers , ICRP 

Publication 68, Annals of the ICRP, Volume 24 Pergamon Press, Oxford, England.

International Commission on Radiological Protection (ICRP). 1996. Conversion 

Coefficients for use in Radiological Protection Against External Radiation. ICRP 

Publication 74. Oxford Pergamon Press.

ICRU, (1993) Quantities and Units in Radiation Protection Dosimetry, ICRU Report No. 

51, International Commission on Radiological Units and Measurements (ICRU), 

Bethesda, Maryland

ICRU, (1997), Dose and Volume Specification for Reporting Interstitial Therapy, ICRU 

Report No. 58,International Commission on Radiological Units and Measurements 

(ICRU), Bethesda, Maryland

Kocher, D. C., Apostoaei, A. I. and Hoffman, F. O. (2002) Radiation Effectiveness 

Factors (REFs) for use in calculating the Probability of Causation of Radiogenic 

Cancers, Health Physics, Vol. 82 (6 (Suppl)), pp. S1888.

National Bureau of Standards (NBS). 1957. Permissible Dose for External Sources of 

Ionizing Radiation. Handbook 59 addendum. US Department of Commerce, 

Washington, D.C.

National Bureau of Standards (NBS). 1954. Permissible Dose for External Sources of 

Ionizing Radiation. Handbook 59. US Department of Commerce, Washington, D.C.

NCRP, (1971) National Council on Radiation Protection and Measurements. Protection 

Against Neutron Radiation, NCRP Report 38, National Council on Radiation Protection 

and Measurements, Bethesda, Maryland.

NCRP, (1996). Screening Models for Releases of Radionuclides to Air, Surface Water, 

and Ground Water National Council on Radiation Protection. NCRP Report No. 123. 

Bethesda, Maryland.

NCRP (1989), Medical X-Ray, Electron Beam and Gamma Ray Protection for Energies 

Up to 50 MeV (Equipment Design, Performance and Use, National Council on Radiation 

Protection, 1989, NCRP Report No. 102, Bethesda, Maryland.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 115 of 188

NCRP, 1987. Exposure of the Population in the United States and Canada from Natural 

Background Radiation. NCRP Report No. 94. National Council on Radiological Protection and 

Measurements, Bethesda, Maryland.

National Research Council (NRC). 1989. Film Badge Dosimetry in Atomospheric 

Nuclear Tests. National Academy Press, Washington, D.C..

Nichols, L. L., G.W.R. Endres, D.B. Shipler, E.E. Oscarson and L.L. Crass. 1972. 

Hanford Multipurpose TL Dosimeter Field Tests and Evaluation. BNWL-B-127, Pacific 

Northwest National Laboratory, Richland, WA.

NIOSH (1993) Epidemiologic use of Nondetectable Values in Radiation Exposure 

Measurements National Institute for Occupational Safety and Health NIOSH Research 

Issues Workshop, September 9-10, 1993, Cincinnati, OH.

NIOSH (2002a), External Dose Reconstruction Implementation Guidelines Rev 1, 

OCAS-IG-001, National Institute of Occupational Safety and Health, Office of 

Compensation Analysis and Support, Cincinnati, OH

NIOSH (2002b), Internal Dose Reconstruction Implementation Guildelines Rev0, 

OCAS-IG-002, National Institute of Occupational Safety and Health (NIOSH, Cincinnati, 

OH

National Research Council (NRC). 1989. Film Badge Dosimetry in Atomospheric 

Nuclear Tests. National Academy Press, Washington, D.C. 

Pardue, L. A., N. Goldstein and E. O. Wollan. 1944. ”Photographic Film As a Pocket 

Radiation Dosimeter. CH-1553-A-2223, Metallurgical Laboratory, Chicago, IL. 

PNNL, 1996, Multimedia Environmental Pollutant Assessment System (MEPAS) 

Application Guidance: Guidance for Evaluating MEPAS Parameters for Version 4.0, 

Pacific Northwest Laboratory, Richland, Washington.

Rabovsky, J.L., C.R. Jones and H.J. Pettengill. 1991. Eleventh DOE Workshop 

on Personnel Neutron Dosimetry. CONF-9106235. Workshop held June 3-7, 

1991, U. S. Department of Energy, Washington, D.C.

Rollins, 2003 – Calculated Air Concentrations for Radionuclides Important for Onsite 

Dose Reconstruction at the Savannah River Site, Dade Moeller & Associates, Augusta, 

GA.

Savannah River Site (SRS). 1993. External Dosimetry Technical Basis Manual. 

WSRC-IM-92-101, dated January 15, 1993. Westinghouse Savannah River Company, 

Aiken, SC 29808.

Savannah River Internal Dosimetry Technical Basis Manual, WSRC-IM-90-139, Rev. 8, 

December 31, 2001

Savannah River Internal Dosimetry Technical Basis Manual, (WSRC-IM-90-139, Rev. 1, 

1990.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 116 of 188

Savannah River Site (SRS). 1993. External Dosimetry Technical Basis Manual. 

WSRC-IM-92-101, dated January 15, 1993. Westinghouse Savannah River Company, 

Aiken, SC 29808.

Scherpelz, R.I., J.J. Fix, and B.A. Rathbone. 2000. Validation of Hanford Personnel and 

Extremity Dosimeters in Plutonium Environments. PNNL-13136. January, 2000. Pacific 

Northwest National Laboratory, Richland, WA 99352.

Singh, L.P. (2002), Letter to Kathryn Robertson-Demers, May 13, 2002, Support of the 

NIOSH Leukemia Case-Control Study.

Seibert, J. A, Barnes, G. T., and Gould, R. G., (1991), Specification, Acceptance Testing 

and Quality Control of Diagnostic X –Ray Imaging Equipment, Medical Physics 

monograph No. 20, American Association of Physics in Medicine, American Institute of 

Physics. 

Taylor, G. A., Crase, K. W., LaBone Thomas R. and Wilkie, W. H., (1995) A History of 

Personnel Radiation Dosimetry at the Savannah River Site, WSRC-RP-95-234, 

Westinghouse Savannah River Company, Aiken, South Carolina.

Thierry-Chef, I., F. Pernicka, M. Marshall, E. Cardis and P. Andreo. 2002. Study of a 

Selection of 10 Historical Types of Dosemeter: Variation of the Response to Hp(10) with 

Photon Energy and Geometry of Exposure. Radiat Prot Dos, 102(2): 101-113.

Thorburn, R.C., May 31, 1950. "Fission Product Analysis of Urine," HW-18320, Health 

Instrument Division, Hanford Works, Richland, WA.

Till, J. E., H. R. Meyer, 1983, Radiological Assessment, Division of Systems Integration, 

Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, 

Washington, D.C.

Till, et al., April 30, 2001, Savannah River Site Environmental Dose Reconstruction 

Project, Phase II: Source Term Calculation and Ingestion Pathway Data Retrieval 

Evaluation of Materials Released from the Savannah River Site

Vallario, E. J, D.E. Hankins and C.M. Unruh. 1969. AEC Workshop on Personnel 

Neutron Dosimetry. BNWL-1340. Battelle, Pacific Northwest Laboratory, Richland, WA.

Vallario, EJ, D.E. Hankins and C.M. Unruh. 1971. Second AEC Workshop on 

Personnel Neutron Dosimetry. BNWL-1616. Battelle, Pacific Northwest Laboratory, 

Richland, WA.

Weber, A. H., Buckley, R. L., and Parker, M.J., 2001, The Creation of an Historical 

Meteorological Database for Dose Reconstruction (U). WSRC-TR-2001-00275, 

Savannah River Technology Center and Oregon State University, Aiken, SC.

Watson, Jr., J.E., J.L. Wood, W.G. Tankersley and C.M. West. 1994. Estimation of 

Radiation Doses for Workers without Monitoring data for Retrospective Epidemiologic 

Studies. Health Phys. 67(4):402-405.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 117 of 188

Westinghouse Savannah River Company (WSRC). 2003. Savannah River Site External 

Dosimetry Technical Basis Manual. WSRC-IM-92-101. Aiken, SC 29808.

Wilson, R.H., J.J. Fix, W. V. Baumgartner, and L. L. Nichols. 1990. Description and 

Evaluation of the Hanford Personnel Dosimeter Program From 1944 Through 1989. 

PNL-7447. Pacific Northwest Laboratory, Richland, WA.

Savannah River Site Environmental Reports

Ashley, C. 1964. Effect of the Savannah River Plant on Environmental Radioactivity 

Semiannual Report, January through June 1964. DPSPU-64-30-2. 

Ashley, C. 1965. Environmental Monitoring at the Savannah River Plant, Annual Report–

1964. DPST-65-302.

Ashley, C. 1966. Environmental Monitoring at the Savannah River Plant, Annual Report 

1965. DPST-66-302. 

Ashley, C. 1967. Environmental Monitoring at the Savannah River Plant, Annual Report–

1966. DPST-67-302. 

Ashley, C. 1968. Environmental Monitoring at the Savannah River Plant, Annual Report–

1967. DPST-68-302. 

Ashley, C. 1969. Environmental Monitoring at the Savannah River Plant, Annual Report–

1968. DPST-69-302. 

Ashley, C. 1970. Environmental Monitoring at the Savannah River Plant, Annual Report–

1969. DPST-70-302. Ashley, C. 1971. Environmental Monitoring at the Savannah River 

Plant, Annual Report–1970. DPST-71-302. 

Ashley, C. 1972. Environmental Monitoring at the Savannah River Plant, Annual Report–

1971. DPST-72-302. 

Ashley, C. and C.C. Zeigler. 1973. Environmental Monitoring at the Savannah River 

Plant, Annual Report–1972. DPSPU-73-302. 

Ashley, C. and C.C. Zeigler. 1974. Environmental Monitoring at the Savannah River 

Plant, Annual Report–1973. DPSPU-74-302. 

Ashley, C. and C.C. Zeigler. 1975. Environmental Monitoring at the Savannah River 

Plant, Annual Report–1974. DPSPU-75-302. 

Ashley, C. and C.C. Zeigler. 1976. Environmental Monitoring at the Savannah River 

Plant, Annual Report–1975. DPSPU-76-302. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 118 of 188

Ashley, C. and C.C. Zeigler. 1978a. Environmental Monitoring at the Savannah River 

Plant, Annual Report–1976. DPSPU-77-302. 

Ashley, C. and C.C. Zeigler. 1978b. Environmental Monitoring at the Savannah River 

Plant, Annual Report–1977. DPSPU-78-302. 

Ashley, C. and C.C. Zeigler. 1981. Environmental Monitoring at the Savannah River 

Plant, Annual Report–1978. DPSPU-79-302. 

Ashley, C. and C.C. Zeigler. 1982. Environmental Monitoring at the Savannah River 

Plant, Annual Report–1979. DPSPU-80-302. 

Ashley, C. and C.C. Zeigler. 1983. Environmental Monitoring at the Savannah River 

Plant, Annual Report 1980. DPSPU-81-302. Health Protection Department, Savannah 

River Plant. Ashley, C. and C.C. Zeigler. 1984. Environmental Monitoring at the 

Savannah River Plant, Annual Report 1981. DPSPU-82-302. Health Protection 

Department, Savannah River Plant, Du Pont. 

Ashley, C., P.C. Padezanin, and C.C. Zeigler. 1984. Environmental Monitoring at the 

Savannah River Plant, Annual Report 1982. DPSPU-83-302. Health Protection 

Department, Savannah River Plant, Du Pont. 

Cummins, C.L., D.K. Martin, and J.L. Todd. 1990. Savannah River Site Environmental 

Report for 1989. Annual Report for 1989. Vol. I and II. WSRC-IM-90-60. Westinghouse 

Savannah River Company. 

Davis, H.A., D.K. Martin, and J.L. Todd. 1989. Savannah River Site Environmental 

Report for 1988. Volume I, Text, Volume II, Figures and Data Tables. WSRC-RP-89-59-

1. Westinghouse Savannah River Company. 

Du Pont. 1959c. Health Physics Regional Monitoring Semiannual Report, January 

through June 1959. DPSPU-59-11-30. 

Du Pont. 1960a. Health Physics Regional Monitoring Semiannual Report, July through 

December 1959. DPSP-60-11-9. 

Du Pont. 1960b. Health Physics Regional Monitoring Semiannual Report, January 

through June 1960. DPSP-60-25-26. 

Du Pont. 1961a. Health Physics Regional Monitoring Semiannual Report, July through 

December 1960. DPSP-61-25-4. 

Du Pont. 1961b. Health Physics Regional Monitoring Semiannual Report–January 

through June 1961. DPSP-62-25-2. 

Du Pont. 1962a. Health Physics Regional Monitoring Semiannual Report, July through 

December 1961. DPSP-62-25-9. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 119 of 188

Du Pont. 1962b. Effect of the Savannah River Plant on Environmental Radioactivity 

Semiannual Report, January through June 1962. DPSPU-62-30-24. 

Du Pont. 1963a. Effect of the Savannah River Plant on Environmental Radioactivity 

Semiannual Report, July through December 1962. DPSPU-63-30-12. 

Du Pont. 1963b. Effect of the Savannah River Plant on Environmental Radioactivity 

Semiannual Report, January through June 1963. DPSPU-63-30-32. 

Du Pont. 1963c. Effect of the Savannah River Plant on Environmental Radioactivity 

Semiannual Report, July through December 1963. DPSPU-64-30-1. 

Du Pont. 1963d. Health Physics Environmental Monitoring Semiannual Report, January 

through June 1962. DPSP-63-25-3. 

Du Pont. 1963e. Health Physics Regional Monitoring Semiannual Report, July through 

December 1962. DPSP-63-25-10. 

Du Pont. 1964. Health Physics Regional Monitoring Annual Report 1963. DPSPU-64-11-

12. Du Pont. 1965a. Effect of the Savannah River Plant on Environmental Radioactivity. 

Semiannual Report, July through December 1964. DPSPU-65-30-1. 

Du Pont. 1965b. Effect of the Savannah River Plant on Environmental Radioactivity. 

Semiannual Report, January through June 1965. DPST-65-30-2. 

Du Pont. 1966a. Effect of the Savannah River Plant on Environmental Radioactivity. 

Semiannual Report, July through December 1965. DPST-66-30-1. 

Du Pont. 1966b. Effect of the Savannah River Plant on Environmental Radioactivity. 

Semiannual Report, January through June 1966. DPST-66-30-2. 

Du Pont. 1967a. Effect of the Savannah River Plant on Environmental Radioactivity. 

Semiannual Report, July through December 1966. DPST-67-30-1. 

Du Pont. 1967b. Effect of the Savannah River Plant on Environmental Radioactivity. 

Semiannual Report, January through June 1967. 

Du Pont. 1968a. Effect of the Savannah River Plant on Environmental Radioactivity. 

Semiannual Report, July through December 1967. DPST-68-30-1. 

Du Pont. 1968b. Effect of the Savannah River Plant on Environmental Radioactivity. 

Semiannual Report, January through June, 1968. DPST-68-30-2. 

Du Pont. 1968c. Effect of the Savannah River Plant on Environmental Radioactivity. 

Semiannual Report, July through December 1968. DPST-69-30-1. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 120 of 188

Du Pont. 1969. Effect of the Savannah River Plant on Environmental Radioactivity. 

Semiannual Report, January through June 1969. DPST-69-30-2. 

Du Pont. 1970a. Effect of the Savannah River Plant on Environmental Radioactivity. 

Semiannual Report, July through December 1969. DPST-70-30-1. 

Du Pont. 1970b. Effect of the Savannah River Plant on Environmental Radioactivity. 

Semiannual Report, January through June 1970. DPST-70-30-2. 

Du Pont. 1971a. Effect of the Savannah River Plant on Environmental Radioactivity. 

Semiannual Report, July through December 1970. DPST-71-30-1. 

Du Pont. 1971b. Effect of the Savannah River Plant on Environmental Radioactivity. 

Semiannual Report, January through June 1971. 

Du Pont. 1972. Environmental Monitoring in the Vicinity of the Savannah River Plant. 

Annual Report, 1971. DPSPU-72-30-1. 

Du Pont. 1973. Environmental Monitoring in the Vicinity of the Savannah River Plant. 

Annual Report, 1972. DPSPU-73-30-1. 

Du Pont. 1974. Environmental Monitoring in the Vicinity of the Savannah River Plant. 

Annual Report, 1973. DPSPU-74-30-1. 

Du Pont. 1975. Environmental Monitoring in the Vicinity of the Savannah River Plant. 

Annual Report, 1974. DPSPU-75-30-1. 

Du Pont. 1977. Environmental Monitoring in the Vicinity of the Savannah River Plant. 

Annual Report, 1976. DPSPU-77-30-1. 

Du Pont. 1978. Environmental Monitoring in the Vicinity of the Savannah River Plant. 

Annual Report, 1977. DPSPU-78-30-1. 

Du Pont. 1979. Environmental Monitoring in the Vicinity of the Savannah River Plant. 

Annual Report , 1978. DPSPU-79-30-1. 

Du Pont. 1980. Environmental Monitoring in the Vicinity of the Savannah River Plant. 

Annual Report, 1979. DPSPU-80-30-1. 

Du Pont. 1981. Environmental Monitoring in the Vicinity of the Savannah River Plant. 

Annual Report, 1980. DPSPU-81-30-1. 

Du Pont. 1982. Environmental Monitoring in the Vicinity of the Savannah River Plant. 

Annual Report, 1981. DPSPU-82-30-1. 

Du Pont. 1983. Environmental Monitoring in the Vicinity of the Savannah River Plant. 

Annual Report, 1982. DPSPU-83-30-1. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 121 of 188

Du Pont. 1984. Environmental Monitoring in the Vicinity of the Savannah River Plant. 

Annual Report, 1983. DPSPU-84-30-1. 

Harvey, R.S., J.H. Horton, and H.G. Mealing, Jr. 1959a. Health Physics Regional 

Monitoring Semiannual Report, January through June 1958. DPSP-58-25-38. 

Harvey, R.S., J.H. Horton, and H.G. Mealing, Jr. 1959b. Health Physics Regional 

Monitoring Semiannual Report, July through December 1958. DPSPU-59-11-23. 

Horton, J.H. 1954. Radioactivity in the Environs of the Savannah River Plant, January to 

July 1954. DP-92. 

Horton, J.H. 1955. Semi-Annual Progress Report, July through December 1954. DPSP55-25-34. 

Horton, J.H. and H.G. Mealing, Jr. 1956. Health Physics Regional Monitoring 

Semiannual Report, July through December 1955. DPSP-56-25-54. 

Mealing, Jr., H.G. 1957a. Health Physics Regional Monitoring Semiannual Report, 

January through June 1956. DPSP-57-25-4.

Mealing, Jr., H.G. 1957b. Health Physics Regional Monitoring Semiannual Report, July 

through December 1956. DPSP-57-25-15. 

Mealing, H.G., Jr., and J.H. Horton. 1957. Health Physics Regional Monitoring 

Semiannual Report, January through June 1957. DPSP-57-25-43. 

Mealing, H.G.,Jr., R.S. Harvey, and J.H. Horton. 1958. Health Physics Regional 

Monitoring Semiannual Report, July through December 1957. DPSP-58-25-17. 

Mikol, S.C., L.T. Burckhalter, J.L. Todd, and D.K. Martin. 1988. Savannah River Plant 

Environmental Report, Annual Report for 1987. Volume I, Text. DPSPU-88-30-1. Health 

Protection Department, Savannah River Plant. 

Zeigler, C.C., I.B. Lawrimore, and W.E. O’Rear. 1985. Environmental Monitoring at the 

Savannah River Plant, Annual Report 1984. DPSPU-85-302. Health Protection 

Department, Savannah River Plant. 

Zeigler, C.C., I.B. Lawrimore, E.M Heath, and J.E. Till. 1986a. Savannah River Plant 

Environmental Report, Annual Report for 1985. Volume I, Text. DPSPU-86-30-1. Health 

Protection Department, Savannah River Plant. 

Zeigler, C.C., I.B. Lawrimore, and E.M. Heath. 1986b. Savannah River Plant 

Environmental Report for 1985. Volume II, Figures and Data Tables. DPSPU-86-30-1. 

Savannah River Plant. 

Zeigler, C.C., E.M Heath, L.B. Taus, J.L. Todd, and J.E. Till. 1987. Savannah River Plant 

Environmental Report, Annual Report for 1986. Volume I, Text, Volume II, Figures and 

Data Tables. DPSPU-87-30-1. Health Protection Department, Savannah River Plant. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 122 of 188
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 123 of 188

ATTACHMENT A 

DESCRIPTION OF FACILITIES AND PROCESSES

The primary operational areas and production facilities are discussed below.

A-1 Reactors – Five heavy-water reactors designated C, K, L, P, and R were constructed at 

SRS in the early 1950s, located in C Area, K Area, L Area, P Area, and R Area, 

respectively. The years of operation and radionuclides are shown in Table A-1.1 and A1.2 below.

Table A-1.1 Reactors Years of Operation

Reactor C K L P R

Years of 

Operation

1955 to1985 1954 to1988 1954 to1968

1985 to1988

1954 to1988 12/28/1953 to 

06/1964

Table A-1.2 Radionuclides of Concern for All Reactors (a)

241Am 60Co 131I 103Ru

140Ba 51Cr 133I 106Ru

141Ce 134Cs 140La 90Sr / 90Y

143Ce 137Cs 54Mn 91Sr

144Ce 152Eu 95Nb 92Sr

242Cm 154Eu 239Np 234U

244Cm 59Fe 238Pu 65Zn

58Co 3

H b 239Pu 95Zr

(a) The composition and relative abundance are not known.

 (b) Tritium is assumed to produce most of the personnel exposure because of the large quantities 

present.

A-2 F-Area A-Line Facility – The A-Line facility accepts dilute, depleted uranyl nitrate 

solution from the 221-F Facility and converts it into uranium trioxide powder. Dilute 

uranyl nitrate solution is sent to A-Line from the solvent extraction process in the 221-F 

Facility. The A-Line facility also has a dissolver to prepare a uranyl nitrate solution from 

the uranium trioxide product and nitric acid when it is required as a process feed stream 

back into the solvent extraction process in the 221-F Facility. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 124 of 188

Table A-2 F Area A-Line Facility

Facility/Process Conversion of Dilute Depleted Uranyl Nitrate Solution (A-Line Feed) to 

Uranium Trioxide Powder (a) (b)

Years of Operation 1954-1987; 1993-1997

Radionuclides

Of Concern 

Absorption 

Types)

Activity Fraction Significant to External 

Exposure

239Pu M 5.559E-03 X

235U F(c) 8.022E-04 X

236U F(c) 8.179E-04 X

238U F(c) 7.440E-01 X

141Ce M 1.305E-02 X

144Ce M 6.079E-02 X

103Ru S(d) 5.969E-02 X

106Ru S(d) 6.111E-02 X

95Nb S(d) 5.418E-02 X

(a) Dosimetrically, the feed material for A-Line is 238U and 239Pu. 

(b) For purpose of calculating internal dose, specific activity of the uranium is approximately 0.4 nCi/mg.

(c) Uranyl nitrate is type F; a solubility study on UO3 powder at Hanford showed it to be mostly type F with a 

small component of type M; assume 80% F, 20% M.

(d) In the nitrate solution the ruthenium would be type F; after the oxidation process or after years of aging 

as contamination, the ruthenium would be type S. Type S will maximize the lung dose; type F maximizes 

the dose to all other tissues. Same statements apply to niobium except it is type M in the nitrate form.

A-3 221-F B-Line Facility – The F B-Line Facility, located in the 200-F Separations Area, is 

used to convert plutonium nitrate into plutonium metal or plutonium oxide and can 

recover plutonium from on-site and off-site scrap. 

Table A-3 221-F B-Line Facility

Facility/Process Conversion of plutonium nitrate into plutonium metal or plutonium oxide

Years of Operation 1954-1989

6 % Mixture (a) (b) 12 % Mixture (a) (c)

Radionuclides

of Concern 

Absorption 

Types Activity Fraction Activity Fraction

Significant to 

External Exposure

238

Pu M-S(d) 4.800E-02 2.660E-01 X 

239Pu M-S(d) 7.760E-01 4.660E-01 X

240Pu M-S(d) 1.760E-01 2.680E-01 X

(a) The percent refers to the mass percent of 240Pu. 

(b) The ratio of 241Pu to alpha Pu is 8 to 1.

(c) The ratio of 241Pu to alpha Pu is 23 to 1.

(d) Pu nitrate is type M; Pu oxides are type S.

A-4 221-F Canyon Facility – The 221-F Canyon Facility consists of two parallel canyons 

that constitute the process area. These canyons contain the high-activity and low-activity 

materials and are referred to as the hot and warm canyons respectively. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 125 of 188

Table A-4.1 221-F Canyon Facility

Facility/Process Head End Stream (Stream 1) Recovery of 239Pu, 237Np and 238U (a) 

Years of Operation 11/1954-1989; 1993-2002

Radionuclides of 

Concern Absorption Types Activity Fraction

Significant to

External Exposure (b)

239Pu M 1.191E-03 X

144Ce M 4.765E-01 X

137Cs F 2.621E-02 X

106Ru F 1.588E-01 X

95Zr M 3.177E-01 X

90Sr F 1.959E-02 X

(a) The function of the 221-F Separation Facility is to isolate and purify plutonium, uranium, and 

neptunium from irradiated uranium. Fission products will also be present and, depending on the age of the 

material, the radionuclides of concern and the fraction of the activity of the materials will change. The 

majority of the internal dose from the Head End stream comes from plutonium, cerium, and ruthenium.

(b) The majority of the external dose from the Head End stream comes from cerium, zirconium, and 

ruthenium.

Table A-4.2 221-F Canyon Facility

Facility/Process Second Uranium Cycle (Stream 2) (a)

Years of Operation 11/1954-1989; 1993-2002

Radionuclides of 

Concern Absorption Types Activity Fraction

Significant to

External Exposure (b) 

239Pu M 4.226E-03

235U M 7.433E-04

238U M 6.893E-01 X

144Ce M 5.633E-02 X

137Cs F 3.133E-03 X

106Ru F 5.662E-02 X

95Zr M 1.873E-01 X

90Sr F 2.332E-03 X

(a) The function of the 221-F Separation Facility is to isolate and purify plutonium, uranium, and neptunium 

from irradiated uranium. Fission products will also be present and, depending on the age of the material, the 

radionuclides of concern and the fraction of the activity of the materials will change. The majority of the 

internal dose from the second uranium cycle (Stream 2) comes from 238U and 239Pu as indicated above. It 

can be assumed this stream consists of depleted uranium with a specific activity of 0.4 nCi/mg and 239Pu (La 

Bone, 1996).

(b) The majority of the external shallow dose from the second uranium cycle (Stream 2) comes from 238U 

which feeds 234mPa, and zirconium which emits both photons and betas.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 126 of 188

Table A-4.3 221-F Canyon Facility

Facility/Process Second Cycle Plutonium Stream (Stream 3) (a)

Years of Operation 11/1954-1989; 1993-2002

Radionuclides of 

Concern Absorption Types Activity Fraction

Significant to 

External Exposure (b)

238Pu M 2.045E-04

239Pu M 6.075E-02

241Pu M 9.194E-01 X

144Ce M 2.424E-03 X

106Ru F 9.204E-04 X

95Zr M 1.616E-02 X 

(a) The function of the 221-F Separation Facility is to isolate and purify plutonium, uranium, and neptunium 

from irradiated uranium. Fission products will also be present, and, depending on the age of the material, the 

radionuclides of concern and the fraction of the activity of the materials will change. 

(b) The majority of the external dose comes from 241Pu and 95Zr. 

A-5 New Special Recovery Facility – Scrap plutonium from offsite sources and from FBLine facility is processed through the New Special Recovery (NSR) Facility. The 

processed plutonium is conditioned so that it can be blended back into the plutonium 

nitrate feed stream to the FB-Line process.

Table A-5 New Special Recovery (NSR) Facility

Facility/Process New Special Recovery (Pu Recovery)

Years of Operation

Radionuclides of 

Concern 

Absorption Types Activity Fraction Significant to External 

Exposure (b)

241Am M

239Pu M To be determined

240Pu M

235U F

236U F

Examples of Feed Material for NSR Facility

Percent of Isotopic Content

Material 239Pu 240Pu 241Am Other

Fe, Ni, Cu, U

Impure Plutonium Oxide 87 to 94 6 to 12 0 to 1 -

Plutonium Metal 88 to 94 6 to 12 - -

Unirradiated Reactor Cores - 4 to 42 0 to 10 60 to 80 U

Scrap MgO Crucibles 88 to 94 6 to 12 0 to 1 < 90 MgO

Mixed Oxides and Metals (c) (c) - 3 Mo

ZPPR Pu and U Oxide Plates 2 to 88 6 to 18 - 78 U, 2 Mo

(a) The majority of the internal dose is delivered by 239Pu, 238Pu, 242Am, 235U, and 236U (assuming enriched 

uranium).

(b) The radionuclide that delivers the majority of the external dose is 241Am. Some neutron dose will also be 

present due to spontaneous fission and an (alpha, n) reaction with fluoride, if present.

(c) Trace plutonium in uranium oxides
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 127 of 188

A-6 F-Area Outside Facilities – The F-Area Outside Facilities provide general support, 

principally to the processing of irradiated fuels and targets in Building 221-F. The term 

“Outside Facilities” is used to describe a wide variety of processes, utilities, and services 

that are ancillary to the primary 200-F Area operations. 

Table A-6.1 F-Area Outside Facilities

Facility/Process Acid Recovery Unit (ARU)

Years of Operation 1954-1989

Radionuclides

of Concern Absorption Types Activity Fraction

Significant to External 

Exposure

234U M 7.040E-11

235U M 1.012E-09

236U M 1.008E-09

238U M 1.008E-07

103Ru F 6.950E-02 X

106Ru F 1.023E-01 X

95Zr M 6.176E-01 X

95Nb M 2.107E-01 X

Table A-6.2 F Area Outside Facilities

Facility/Process Water Handling System

Years of Operation 1954-1989

Radionuclides of 

Concern 

Absorption Types Activity Fraction Significant to External 

Exposure

103Ru S 1.417E-01 X

106Ru S 4.815E-01 X

95Zr S 1.298E-01 X 

95Nb S 2.470E-01 X
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 128 of 188

Table A-6.3 F Area Outside Facilities

Facility/Process General Purpose Evaporators

Years of Operation 1954-1989

Radionuclides of 

Concern Absorption Types Activity Fraction

Significant to External 

Exposure (b)

238Pu S 9.626E-05

239Pu S 2.903E-02

241Pu S 3.913E-01 X

234U S 2.911E-08

235U S 4.813E-07

236U S 4.343E-07

238U S 9.343E-05 X

103Ru S 4.852E-02 X

106Ru S 3.369E-01 X

95Zr S 9.743E-02 X

95Nb S 9.665E-02 X

(a) The radionuclide of concern for internal exposure for the GP Evaporators, Segregated Plutonium 

Solvent, and Laboratory Waste is plutonium. Type M (type F for ruthenium) should be used for nitrate form 

or from solvent extraction. 

(b) The radionuclides of external dosimetric concern for the GP Evaporators, Segregated Plutonium Solvent, 

and Laboratory Waste are as indicated above. 238U feeds 234Pa with a high energy beta which is also of 

concern.

Table A-6.4 F Area Outside Facilities

Facility/Process Segregated Plutonium Solvent

Years of Operation 1954-1989

Radionuclides

of Concern Absorption Types Activity Fraction

Significant to External 

Exposure (b)

238Pu M 2.068E-04

239Pu M 5.941E-02

241Pu M 8.701E-01 X

103Ru F 2.418E-03 X

106Ru F 5.941E-02 X

95Zr M 5.941E-03 X

95Nb M 2.532E-03 X

(a) The radionuclide of concern for internal exposure at the GP Evaporators, Segregated Plutonium Solvent, 

and Laboratory Waste is plutonium.

(b) The radionuclides of external dosimetric concern for the GP Evaporators, Segregated Plutonium Solvent, 

and Laboratory Waste are as indicated above. 238U feeds 234Pa with a high energy beta which is also of 

concern.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 129 of 188

Table A-6.5 F Area Outside Facilities Table 

Facility/Process Laboratory Waste

Years of Operation 1954-1989

Radionuclides

of Concern 

Absorption 

Types(a) Activity Fraction

Significant to External 

Exposure (b) 

238Pu S 6.150E-01

239Pu S 1.921E-02

241Pu S 3.456E-01 X

234U S 4.803E-08

235U S 7.497E-07

236U S 7.209E-07

238U S 7.497E-05 X

141Ce M 1.369E-04 X

144Ce M 9.892E-03 X

134Cs F 1.822E-04 X

137Cs F 1.590E-03 X

103Ru S 1.369E-04 X

106Ru S 8.645E-04 X

95Zr M 6.150E-03 X

95Nb S 1.237E-03 X

(a) The radionuclide of concern for internal exposure at the GP Evaporators, Segregated Plutonium Solvent, 

and Laboratory Waste is plutonium. Various solubility classes are possible; those listed can be used as 

defaults. Nitrate forms would have been more soluble.

(b) The radionuclides of external dosimetric concern for the GP Evaporators, Segregated Plutonium Solvent, 

and Laboratory Waste are as indicated above. 238U feeds 234Pa with a high energy beta which is also of 

concern.

Table A-6.6 F Area Outside Facilities

Facility/Process Segregated Uranium Solvent

Years of Operation 1954-1989

Radionuclides of

Concern Absorption Types Activity Fraction

Significant to External 

Exposure

234U M 3.371E-07

235U M 5.941E-06

236U M 5.775E-06

238U M 5.775E-04 X

103Ru F 1.491E-01 X

106Ru F 6.417E-01 X

95Zr M 1.491E-01 X

95Nb S 5.941E-02 X
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 130 of 188

A-7 238PuO2 Fuel Form Facility (PuFF) and the 238PuO2 Experimental Facility (PEF) – 

The primary function of the PuFF facility is to produce encapsulated plutonium oxide fuel forms. 

The conversion of plutonium oxide powder into dense PuO2 fuel forms is conducted in the PEF.

Table A-7 238PuO2 Fuel Form Facility (PuFF) and 238PuO2 Experimental Facility (PEF)

Facility/Process Production of encapsulated plutonium oxide fuel forms

Years of Operation 1978 -??

Radionuclides 

of Concern a

Absorption 

Types Activity Fraction Specific Activity (Ci/g)

Significant to 

External Exposure

241Am S(b) 3.067E-04 3.21 X

238Pu S 8.535E-01 1.66E+1 X

239Pu S 1.411E-01 6.13E-2

240Pu S 2.27E-1

241Pu S 4.089E-03 1.12E+2

242Pu S 3.85E-3

237Np M 5.111E-04 6.87E-4 X

232Th M 5.111E-04 1.11E-7 X

Particle Size 

Mass Median Diameter

Ball-milled Powder 2 μm

Feed Powder 5 μm

(a) The majority of the dose comes from plutonium.

(b) Type M if mostly pure; type S if a contaminant in a plutonium matrix.

A-8 235-F Vaults – The 235-F facility receives plutonium and plutonium dioxide and 

fabricates it into a variety of items used in the nuclear program. There are three storage 

vaults located in the 235-F facility, the Plutonium Oxide Vault, Finished Product Vault, 

and Scrap Vault.

Table A-8 235-F Vaults

Facility/Process Plutonium Oxide Vault; Finished Product Vault, Scrap Vault (a)

Years of Operation

Radionuclides

of concern Absorption Types Activity Fraction

Significant to External 

Exposure

235U S

238Pu S

239Pu S

237Np M

(a) Detailed information on the composition of the radionuclides stored and found in these vaults is not 

available; therefore, they are considered to be pure.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 131 of 188

A-9 772-F and 772-1F Production Control Laboratories – The production control 

laboratories provide analytical support for the 200-F and 200-H separations processes. 

Table A-9 772-F and 772-1F Production Control Laboratories

Facility/Process 772-F & 772-1F Production Control Laboratories (a)

Years of Operation

Radionuclides 

of Concern Absorption Types Activity Fraction

Significant to External 

Exposure

Pu M

U M

Fission 

Products

(a) Specific fission products or Pu and U information is not specified.

A-10 E-Area Solid Waste Disposal Facility (SWDF) – Solid and liquid low-level waste 

generated at the SRS and the Savannah River Laboratory (SRL) and some waste from 

off-site are stored at the SWDF.

Table A-10 E-Area Solid Waste Disposal Facility (SWDF)

Facility/Process Liquid and Solid Waste Disposal (a)

Years of Operation

Radionuclides 

of Concern Absorption Types Activity Fraction

Significant to External 

Exposure

(a) Due to the diverse nature of the radioactive waste, the radionuclide content of each type of waste and 

absorption types has not been defined. The radionuclides of highest concentration (Ci/vol) are tritium, cobalt, 

plutonium, and curium.

A-11 Plutonium Storage Facility (PSF) – The Plutonium Storage Facility (PSF) is located in 

the 200-F Separations Area and is used to receive, store, monitor, retrieve, and ship 

packaged plutonium, as needed. 

Table A-11 Plutonium Storage Facility (PSF)

Facility/Process Receipt, Store, Monitor, Retrieve and Ship Plutonium 

Years of Operation

Radionuclides 

of Concern 

Absorption 

Types

Activity Fraction (a) (b) (d) 

@ 6 % Pu Button 

Activity Fraction (a) (c) (d) 

@12 % Pu Button

Significant to 

External 

Exposure 

238Pu S 4.800E-02 2.660E-01

239Pu S 7.760E-01 4.660E-01

240Pu S 1.760E-01 2.680E-01

(a) The percent refers to the mass percent of 240Pu.

(b) The ratio of 241Pu to alpha Pu is 8.0 to 1.

(c) The ratio of 241Pu to alpha Pu is 23 to 1.

(d) The activity fractions of alpha-emitting plutonium (alpha-Pu) isotopes are those found in typical 6% or 

12% plutonium buttons from the 221-F B Line.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 132 of 188

A-12 F/H Effluent Treatment Facility (ETF) – The F/H Effluent Treatment Facility (ETF) 

consists of large impermeable storage basins with lift stations located in both F and H 

Areas and a treatment facility, Building 241-84H located in H Area. ETF is designed to 

remove hazardous chemical and radioactive contaminants from the 200-Area liquid 

effluent waste streams and concentrate them for disposal. 

Table A-12 F/H Effluent Treatment Facility (ETF)

Facility/Process Removal and Concentration of Waste from Liquid Waste Streams 

Years of Operation 1988- ??

Radionuclides of 

Concern

Absorption Types Activity Fraction Significant to External

Exposure

238Pu S (Y)(a) 2.855E-04

239Pu S (Y)(a) 2.515E-06

241Pu S (Y)(a) 8.401E-04

144Ce M (W)(a) 4.690E-01 X

134Cs F 1.340E-02 X

137Cs F 2.723E-02 X

106Ru S 7.763E-02 X

95Zr M 2.382E-01 X

89Sr F 1.601E-01 X

90Sr F 1.335E-02 X

(a) Lung Solubility Classes were chosen to maximize the missed committed dose.

A-13 F & H Cooling Water and Retention Basins – The F and H Areas consist of chemical 

separation plants and appropriate support facilities for the production of purified radionuclides, 

principally uranium, plutonium, and tritium. The cooling water and retention basins are support 

facilities for the separations and waste management areas. Low-level radioactive liquid wastes 

from the separations areas are routed to the basins when activity is detected.

Table A-13 F & H Area - Cooling Water and Retention Basins

Facility/Process F- Area Retention 

Basins (a) 

H - Area 

Retention 

Basins (a)

Years of Operation 1955-1989

Radionuclides 

of concern 

Absorption 

Types

Activity Fraction Activity Fraction Significant to External 

Exposure (b) 

238Pu S 7.007E-03 1.403E-02

239Pu S 2.803E-02 1.303E-02

238U S 6.206E-02 9.018E-03 X

144Ce S 1.001E-03 1.403E-02 X

137Cs F 7.047E-01 7.335E-01 X

106Ru S 7.508E-02 2.405E-02 X

90Sr F 1.221E-01 1.924E-01 X

(a) The majority of the annual internal dose from the F-Area and H-Area Basins comes from plutonium and 

uranium.

(b) External exposure is for both F and H Area Retention Basins.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 133 of 188

A-14 F & H Area Tank Farms – High-level liquid radioactive wastes from SRS are received 

and managed in large underground tanks in the waste tank farms in 241-F and 241-H. The two 

farms contain 51 large subsurface tanks and related facilities required for safe handling, 

processing, and temporary retention of liquid wastes. 

Table A-14 F & H Area Tank Farms

Facility/Process F-Area Combined 

Tank Waste

H-Area Combined

Tank Waste

Years of Operation 1955-1989

Radionuclides 

of concern 

Absorption 

Types Activity Fraction Activity Fraction

Significant to 

External Exposure (b)

241Am M 1.319E-04 7.346E-05 X

244Cm M 9.341E-04 6.121E-06

238Pu M 2.198E-05 5.632E-03 X (c) 

239Pu M 1.099E-04 -

241Pu M 1.022E-02 5.693E-03 X

144Ce M 1.538E-01 2.632E-01 X

137Cs F 4.286E-01 3.489E-01 X

106Ru S 1.055E-02 2.142E-02 X

90Sr F 3.956E-01 3.550E-01 X

(a) When unknown or variable lung solubility classes were chosen to maximize the missed dose.

(b) External exposure is for both F and H Area Combined Tank Waste.

(c) For the F-Area Combined Tank Waste, 238Pu is not a radionuclide of concern for external exposure.

A-15 Waste Certification Facility (WCF) – The Waste Certification Facility (WCF), Building 

724-8E, located in the 643-E Solid Waste Disposal Facility, is used to certify and 

package drums of transuranic (TRU) waste. The functions of the WCF are to assay and 

certify the contents of TRU waste drums and to package, load, and ship the drums.

Table A-15 Waste Certification Facility (WCF)

Facility/Process Building 724-8E Certification and Packaging of TRU waste

Years of Operation

Radionuclides 

of concern (a) Absorption Types Activity Fraction

Significant to 

External

Exposure

(a) Radionuclides are considered to be the same as for the Solid Waste Disposal Facility (643-E). The 

radionuclides in the solid waste disposal facility have not been identified. It is assumed the radionuclides are 

tritium, curium, plutonium, and activation products.

A-16 221-H B-Line Facility – The HB Line consists of three separate facilities; the

• Scrap Recovery Facility

• Neptunium Oxide Facility, and 

• Plutonium Oxide Facility.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 134 of 188

Table A-16.1 221-H B-Line Facility

Facility/Process Scrap Recovery 238U, 239U 

Years of Operation

Isotopic composition 

of 238Pu scrap

Isotopic composition 

of 239Pu scrap

Radionuclides 

of concern

Absorption 

Types

Activity Fraction Activity Fraction

Significant to 

External 

Exposure (a)

238Pu S 9.601E-01 - X

239Pu S 1.200E-03 1.003E-01 X

241Pu S 3.870E-02 8.997E-01 X

(a) External exposure is for both 238Pu and 239Pu.

Table A-16.2 221-H B-Line Facility

Facility/Process Scrap Recovery 235U

Years of Operation

Isotopic composition of 235U scrap

Radionuclides

of concern

Absorption 

Types

Activity Fraction Significant to External 

Exposure

233U S 9.556E-01 X

235U S 4.020E-02 X

236U S 3.800E-03 X

238U S 4.000E-04 X 

Table A-16.3 221-H B-Line Facility. Np02 andPu02 Facility

Facility/Process Neptunium Oxide 

Facility

Plutonium Oxide 

Facility

Years of Operation

Isotopic composition 

of NpO2 Material (a)

Isotopic composition 

of PuO2 Material

Radionuclides

of concern 

Absorption 

Types NpO2 Activity 

Fraction

PuO2 Activity Fraction

Significant to 

External 

Exposure (b)

237Np M 1.000E+02 - X

238Pu S - 9.601E-01 X 

239Pu S - 8.992E-04 X

241Pu S - 2.997E-02 X

(a) For purposes of dose calculations the NpO2 is assumed to be 100% 237Np.

(b) External exposure is for both facilities.

A-17 H-Canyon Facility – The H-Canyon building was designed for the separation and 

recovery of 239Pu and 238U from irradiated natural uranium by the Plutonium Uranium 

Extraction (PUREX) (Facility/Process) process. Over the years, other processes have 

been added. These operations include the processing of irradiated enriched uranium to 

recover 235U, the processing of irradiated neptunium targets to separate and recover 

238Pu and 237Np, and the processing of thorium to recover 233U. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 135 of 188

Table A-17.1 H-Canyon Facility 

Facility/Process First Cycle Stream- Uranium Cycle (a)

Years of Operation 1955-1991

Radionuclides 

of concern Absorption Types Activity Fraction

Significant to External 

Exposure

238Pu M 5.719E-04 X

239Pu M 4.981E-06

241Pu M 1.679E-03 X 

237Np M 1.587E-07

234U M 2.398E-06

235U M 3.321E-08

236U M 3.136E-07

238U M 8.671E-10

144Ce M 5.535E-01 X

137Cs F 3.321E-02 X

106Ru F 2.583E-02 X 

95Zr M 2.214E-01 X

89Sr F 1.531E-01 X

90Sr F 1.070E-02 X

(a) The majority of the annual dose comes for the First Cycle Stream comes from Pu and Ce.

Table A-17.2 H-Canyon Facility 

Facility/Process Second Uranium Cycle Stream (a)

Years of Operation 1955-1991

Radionuclides 

of concern Absorption Types Activity Fraction

Significant to External 

Exposure

238Pu M 9.016E-05

239Pu M 7.889E-07

241

Pu M 2.630E-04 X

237Np M 5.071E-04 X

234U M 1.878E-01 X

235U M 2.630E-03 X 

236U M 2.630E-02 X

238U M 6.950E-05

144Ce M 4.320E-01 X

137Cs F 2.442E-02 X 

106Ru F 2.066E-02 X

95Zr M 1.766E-01 X

89Sr F 1.202E-01 X

90Sr F 8.452E-03 X 

(a) The majority of the annual dose from the Second Uranium Cycle Stream comes from uranium. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 136 of 188

Table A-17.3 H-Canyon Facility 

Facility/Process Neptunium Second Cycle (a)

Years of Operation 1955-1991

Radionuclides 

of concern Absorption Types(b) Activity Fraction

Significant to External 

Exposure

238Pu S 3.596E-03 X

239Pu S 3.083E-05

241Pu S 1.028E-02 X

237Np M 4.624E-04 X

234U S 1.284E-06

235U S 1.284E-06

236U S 1.284E-06

144Ce S 1.670E-01 X

137Cs F 1.002E-02 X

106Ru S 1.284E-02 X

95Zr M 6.936E-01 X

89Sr F 9.890E-02 X

90Sr F 3.339E-03 X

(a) The majority of the annual dose for the Second Neptunium Cycle Stream comes from plutonium and 

cerium.

(b) The plutonium would have been type M in the nitrate phase of the process and type S in the oxalate and 

oxide phases. Uranyl nitrate would have been type F; uranium trioxide: 80% F, 20% M.

Table A-17.4 H-Canyon Facility 

Facility/Process Work Stream (a)

Years of Operation 1955-1991

Radionuclides 

of concern Absorption Types Activity Fraction

Significant to External 

Exposure

238Pu S 5.719E-04 X

239Pu S 4.981E-06

241Pu S 1.679E-03 X

237Np M 1.476E-09

234U S 2.214E-10

235U S 3.690E-12

236U S 2.952E-11

238U S 7.380E-14

144Ce S 5.535E-01 X

137Cs F 3.321E-02 X

106Ru S 2.583E-02 X

95Zr M 2.214E-01 X

89Sr F 1.531E-01 X

90Sr F 1.070E-02 X

(a) The majority of the annual dose for the Work Stream comes from plutonium and cerium.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 137 of 188

A-18 211-H Area Outside Facilities – H-Area Outside Facilities include a number of 

processes that support the separations function of the 200-H Area. 

Table A-18.1 H-Area Outside Facilities

Facility/Process A Line (a)

Years of Operation 1955-1991

Radionuclides 

of concern Absorption Types Activity Fraction

Significant to External 

Exposure (b)

234U F 8.055E-01 X

235U F 8.055E-03 X

236U F 1.859E-01 X

238U F 5.164E-04 X

(a) The radionuclide of concern for A-Line is the enriched uranium as indicated.

(b) The major contributor to external dose will be photons and x-rays from 234U, 235U, and 236U.

Table A-18.2 H - Area Outside Facilities

Facility/Process Water Handling System 

Years of Operation 1955-1991

Radionuclides 

of concern Absorption Types(b) Activity Fraction

Significant to External 

Exposure (a)

137Cs F 9.990E-04 X

106Ru S 7.992E-02 X

95Zr M 4.096E-01 X

95Nb S 5.095E-01 X

(a) The radionuclides for the other facilities are 95Nb, 95Zr, 103Ru, 106Ru, and 137Cs. However, the mixture of 

radionuclides for the Water Handling System is considered typical of the other Outside Facilities: Acid 

Recovery Unit, General Purpose Evaporators, Segregated Solvent Facilities, Transfer Tanks, Sump 

Collection Tanks, and Recycle Sumps.

(b) Default type unless material is known to have been in the nitrate form, e.g, in the ARU, in which case Ru 

is type F, Zr is type F, and Nb is type M.

A-19 Receiving Basin for Off-site Fuel (RBOF) and Resin Regeneration Facility (RRF) – 

The RBOF facility handles casks and fuels of various shapes, sizes, and content. Casks 

of spent fuel are received and washed in the cask basin to remove any dirt accumulated 

during transit. The solubility classes in Table 19.1 are SR-1 for 131I and SR-0 for gases 

or vapors that are inert and constitute a possible external exposure from submersion in 

cloud of gas and to internal exposure from gas within the respiratory tract. Class SR-1 

refers to soluble or reactive gases or vapors with consideration given to retention in 

respiratory tract tissues, and to uptake to the systemic circulation, which may be less 

than the 100 % of the inhaled activity (ICRP 68, page9)
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 138 of 188

Table A-19.1 Receiving Basin for Off-site Fuel (RBOF)

Facility/Process Irradiated Fuel Receipt and Storage (a)

Years of Operation

Radionuclides 

of concern Absorption Types Activity Fraction

Significant to External 

Exposure

131I SR-1 1.020E-01 X

85Kr-85 SR-0 8.900E-01 X

133Xe SR-0 5.600E-03 X

(a) The irradiated fuel contains uranium, fission products and activation products. Since the work is 

performed under water, the inert gases that escape the fuel bundles are a primary concern. The distribution 

above is for a Mark 22 Fuel Assembly cooled for 90 days.

Table A-19.2 Resin Regeneration Facility (RRF)

Facility/Process Spent Ion Exchange Resin Regeneration (a)

Years of Operation

Radionuclides 

of concern Absorption Types Activity Fraction

Significant to External 

Exposure

134Cs F 1.500E-01 X

137Cs F 1.500E-01 X

60Co S 7.000E-01 X

(a) The resins contain fission and activation products with the approximate distribution above. Irradiated 

target elements that are cut and cleaned at RRF contain tritium.

A-20 Tritium Facilities – The Tritium Facility includes Building 232-H, 233-H, 234-H, and 

238-H. Irradiated lithium targets are processed in the tritium facilities to recover, purify, 

and package tritium.

Table A-20 Tritium Facilities (a)

Facility/Process Buildings 232-H, 233-H, 234-H, and 238-H

Years of Operation 1952-Present

Radionuclides of 

concern Absorption Types Activity Fraction

Significant to External 

Exposure

Elemental Tritium

Tritiated water

65Zn S

(a) Elemental tritium and tritiated water are assumed to be the primary radionuclides of concern at the tritium 

facilities. In addition, the activation product 65Zn is associated with the processing of lithium targets in 

building 232-H.

A-21 M-Area – Uranium Target Fabrication Facility, 313-M - The Uranium Target Fabrication 

Facility was designed and built to manufacture aluminum clad targets for irradiation in 

the SRS reactors.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 139 of 188

Table A-21 Uranium Target Fabrication Facility, 313 M

Facility/Process Uranium Target Fabrication (a)

Years of Operation 1953-1989

Radionuclides of 

concern Absorption Types(b) Activity Fraction

Significant to External 

Exposure

234U S 8.399E-02 X

235U S 1.450E-02 X

238U S 9.015E-01 X

(a) No information is available on the radionuclide distribution of the 313-M Facility. Therefore the 

distribution has been assumed. The majority of the annual dose from this mixture is delivered by 238U.

(b) Based on solubility studies at similar facilities at Hanford, there may have been a type F component at

about 10-20%. 100% type S should be assumed unless the urinalysis data show otherwise. 

A-22 320-M, 322-M, and 341-M – Target extrusion was conducted in Building 320-M, and 

Building 322-M was a Chemical and Metallurgical Laboratory. The M-Area facilities 

processed uranium, lithium, and aluminum into fuel and target components for the 

nuclear reactors. 

Table A-22 Uranium Processing Facilities 320-M, 322-M, and 341-M (a)

Facility/Process Depleted Uranium (b) Enriched 

Uranium (c)

Years of Operation 1953-1989

Radionuclides 

of concern 

Absorption 

Types(d) Activity Fraction Activity 

Fraction

Significant to 

External Exposure (e) 

234U S 8.399E-02 8.500E-01 X

235U S 1.450E-02 1.100E-02 X

236U S - 1.380E-01 X

238U S 9.015E-01 1.000E-03 X

(a) No information is available on the processes or radionuclide distribution of the 320-M, 322-M, or 341-M 

Facilities, therefore the above distribution has been assumed.

(b) The majority of the annual internal dose from this mixture is delivered by 238U.

(c) The majority of the annual internal dose from this mixture is delivered by 234U.

(d) Based on solubility studies at similar facilities at Hanford, there may have been a type F component at 

about 10-20%. 100% type S should be assumed unless the urinalysis data show otherwise.

(e) The majority of the external shallow dose from this mixture is delivered by 238U which decays to 234Pa 

which emits a high energy beta. External exposure is for both types of uranium.

A-23 Fuel Fabrication Facility, 321-M – The Fuel Fabrication Facility was designed and built 

to manufacture aluminum clad fuel elements for irradiation in the SRS reactors.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 140 of 188

Table A-23 Fuel Fabrication Facility, 321-M (a) 

Facility/Process Fuel Fabrication

Years of Operation 1953-1989

Radionuclide content of 

material being 

processed in 1996

Mass 

Fraction

Absorption 

Types(b) Activity Fraction (C)

Significant to 

External 

Exposure

234U 0.015 S 8.500E-01 X

235U 0.580 S 1.100E-02 X

236U 0.245 S 1.380E-01 X

238U 0.160 S 1.000E-03 X

237Np 1.0 - - X

(a) Plutonium dioxide was processed in 321-M in 1980 and Pu-Al Alloy work was conducted in the 1960s.

(b) Based on solubility studies at similar facilities at Hanford, there may have been a type F component at 

about 10-20%. 100% type S should be assumed unless the urinalysis data show otherwise. 

(c) The majority of the dose from this mixture is delivered by 234U.

A-24 S-Area- Defense Waste Processing Facility (DWPF) – Radioactive wastes are stored 

in existing H-Area waste tank facilities in the form of settled sludge, supernatant liquid, 

and salt cake. The insoluble solids (sludge) undergo aluminum dissolving and washing 

operations in preparation for the feed to DWPF. 

Table A-24.1 S Area Defense Waste Processing Facility (DWPF)

Facility/Process Sludge Slurry Feed Stream (Stream 1) 

Years of Operation 1996-???

Radionuclides 

of concern 

Absorption 

Types

Activity Fraction (a) (b) Solubility (c)

Significant to 

External 

Exposure

244Cm M 1.682E-03 Unknown X

238Pu S 2.326E-02 PuO2, PuO2(NaTi2O5)2, 

Na2PuO2(OH)4 (sol)

X 

241Pu S 2.684E-02 Same as 238Pu X

144Ce S 1.557E-01 Unknown X

137Cs F 2.326E-02 Cs2O, CsOH (sol.) X 

106Ru S 3.578E-02 Na2RuO4,(sol.) X 

90Sr F 7.335E-01 Sr(NaTi2O5)2, Sr(OH)2(sol.), 

SrCO3, SrO

X 

(a) Specific activity 56 Ci/gal.

(b) The majority of the dose from this mixture is delivered by 90Sr and 238Pu.

(c) Chemical Form is insoluble unless otherwise indicated. Pu oxides are type S, but there is some 

indications that pure 238Pu oxides are removed from the lung more rapidly than 239Pu oxides.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 141 of 188

Table A-24.2 S Area Defense waste Processing Facility (DWPF)

Facility/Process Precipitate-Slurry Feed Stream (Stream 201)

Years of Operation 1996-???

Radionuclides 

of concern 

Absorptio

n Types

Activity 

Fraction (a) b) Solubility (c) 

Significant to 

External 

Exposure (d)

244Cm M 2.537E-05 Unknown X

238Pu S 3.171E-04 Same as sludge slurry X 

241Pu S 2.537E-04 Same as sludge slurry X

144Ce S 1.839E-01 X

134Cs F 3.594E-03 CsO, CsOH (sol.), CsB(CH) X

137Cs F 7.611E-01 Same as 134Cs X

106

Ru S 4.228E-02 X

90Sr F 8.457E-03 Sr(NaTi O)2, Sr(OH)(sol.), SrCO X 

(a) Specific activity is 37 Ci/gal.

(b) The majority of the internal dose from this mixture is delivered by 137Cs and 144Ce.

(c) Chemical Form is insoluble unless otherwise indicated.

(d) The majority of the external dose comes from 144Ce, 137Cs, 106Ru, 90Sr, 238Pu, and 234Pu. 

A-25 Z Area – Waste tank supernate (low-level waste) is transferred from H Area to Z Area 

via pipeline. At Z Area, the salt solution (supernate) is mixed with cement, flyash, and 

blast furnace slag, and the resulting grout is pumped to vaults where it sets into a 

saltstone monolith.

Table A-25.1 Z Area

Facility/Process Waste Tank Supernate - Decontaminated Salt Solution

Years of Operation

Radionuclides 

of concern

15 -Year Aged (a)

(nCi/g)

Absorption 

Types Activity Fraction

Significant to External 

Exposure (b)

3

H 15.7

90Sr 0.8 (Including Y) F 1.374E-2 X

Tc-99 53.0

106Ru 60 (including Rh) M 5.888E-01 X

Sb-125 8.0

137Cs 9.4 F 3.925E-01 X

238Pu 0.6 M 4.907E-03 X

239Pu 0.007 M

All TRU 

elements

0.9 M 

(a) The compositions are based on the tetraphenylborate and the sodium titanate (NaTi2O5) precipitation 

processes for removing 137Cs and 90Sr, respectively from the salt solution. Composition varies according to 

tanks.

(b) The majority of the external dose from by the salt solution is delivered by 238Pu, 137Cs, 90Sr, and 106Ru.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 142 of 188

Table A-25.2 Z Area

Facility/Process Saltstone

Years of Operation

Radionuclides 

of concern 

15 -Year Aged (a)

(nCi/g)

Absorption 

Types

Activity 

Fraction

Significant to External 

Exposure (b)

3

H 7.0

90Sr 0.36 (double

for Y)

S 1.401E-02 X 

99Tc 25.0

106Ru 26 (Double Rh) M 5.602E-01 X

125Sb 4.0

137Cs 4.4 F 4.202E-01 X

238Pu 0.3 M 5.602E-03 X

239Pu 0.003

All TRU 

elements

0.4

(a) The compositions are based on the tetra-phenylborate and the sodium titanate (NaTi2O5) precipitation 

processes for removing 137Cs and 90Sr, respectively from the salt solution. Composition varies according to 

tanks.

(b) In the saltstone, the majority of the external dose is delivered by 238Pu, 137Cs, 90Sr, and 106Ru.

A-26 A Area Building 773-A – Building 773-A has many processes including the following: 

Actinide Technology, Analytical Development, Defense Waste Processing Technology, 

Hydrogen Technology, Interim Waste Technology, Environmental Technology, 

Laboratory Services, Materials Technology.

Table A-26 A Area

Years of Operation 1954-1989

Additional documents are needed to review the facilities in the A Area for both internal and external 

radionuclides of concern.

There are many "Sections" and "Technologies" in this area.

No information is currently available

A-27 Buildings 735-A and 735-11A – Buildings 735-A and 735-11A house the Radiological 

and Environmental Science facilities. 

A-28 Building 776-A Liquid Waste Handling Facility – Located in Building 776-A, this 

facility collects the aqueous waste from Building 773-A. Facilities in Building 776-A are 

provided to strain solids from the waste streams, collect waste in batch receiving tanks, 

sample and adjust the pH of the waste tank contents, and transfer the waste to tank 

trailers for disposal. 

A-29 D-Area Heavy Water Production and Reprocessing – A heavy water production plant, 

in D-Area, began operation early in SRS history to concentrate heavy water from 

Savannah River to moderate and cool the site’s reactors.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 143 of 188

ATTACHMENT B 

OCCUPATIONAL MEDICAL DOSE

B.0 SRS Medical X Ray

SRS conducted pre-employment and annual physical examinations as part of their 

occupational health program. These examinations typically included a chest x-ray. For 

some workers, and occupations, chest x-rays could be more frequent. 

B.1 Organ Doses from Medical X-Rays

X-ray organ doses for occupational x-rays at SRS are estimated for all years from 1950 

to present. The schedule for these exams for all SRS employees over this time period is 

shown in Table B-01 and B-02, along with the organ dose information. X-ray organ dose 

estimates were made for Type I equipment (used from 1950 to 1970), Type II equipment 

(used from 1971 to 7/1985), Type III equipment (used from 8/1985 to 5/1999), and Type 

IV equipment (used from 6/1999 to 2002). No collimation was assumed for Type I 

equipment used prior to 1970. For organs outside the chest cavity, other ICRP 34 organ 

dose conversion factors besides PA chest were used to ensure that the organ doses 

reflected their presence in the primary beam. For example, the dose to the ovaries for 

Type I equipment assumes they were in the primary beam, so the air kerma at the skin 

was multiplied by the dose conversion factor for ovaries in the abdomen.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 144 of 188

Table B-01 Doses for Organs Identified in ICRP 34 (1982) Beam Quality for 3.5 mm Al HVL

Time 

Period Frequency

SRS 

Employee

Organ Doses per PA Chest (a) (rem)

Air Kerma 

at Skin 

Entrance Thyroid Ovaries Testes Lungs Breast Embryo

Bone 

Marrow WB

1950 - 0.108 0.107 (c) 0.032 (c) 0.0019 (c) 0.0610 (m) 0.0098 0.028 (c) 0.0158(m) 0.0188 (m)

1970

(Type I)(b) Annual All 0.0650 (f) 0.0152(f) 0.0174 (f)

1971-

7/1985

(Type II) Annual All

0.044 0.0027 0.00014 <0.0000004 0.0249 (m) 0.0040 0.00013 0.0064(m) 0.0076 (m)

 0.0268 (f) 0.0062(f) 0.0071 (f)

8/1985 - 

5/1999

(Type III) Annual All

0.033 0.0020 0.00011 <0.0000003 0.0186 (m) 0.0030 0.00010 0.0048(m) 0.0057 (m)

0.0201 (f) 0.0047(f) 0.0053 (f)

6/1999 -

Annual

Over 50 0.033 0.0020 0.00011 <0.0000003 0.0186 (m) 0.0030 0.00010 0.0048(m) 0.0057 (m)

2003

(Type IV) years old 0.0201 (f) 0.0047(f) 0.0053 (f)

Biennial

40-49 

years old

Every 3rd

year

Under 40 

years old a Organ doses for lateral chests can be estimated at 2.5 times the organ doses for PA chests (Ron Kathren and Vern Shockley)

b

Doses from Type I equipment are estimated at 2.5 times the doses from Type II equipment (Ron Kathren and Vern Shockley)

c No collimation assumed prior to 1970. For organs not in the chest cavity, other ICRP dose conversion factors were used 

(abdomen, skull, and cervical spine).
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 145 of 188

Table B-02 Dose For IREP Organs Not Included in ICRP 34 (1982) Beam Quality for 3.5 mm Al HVL

Time 

Period Frequency

SRS 

Employee

Organ Doses per PA Chest (a) (rem)

Air 

Kerma Thymus

Liver/Gall 

Bladder Eye/Brain

EsophagusStomach UterusUrinary/ 

Bladder

Colon/ 

Rectum

Bone 

Surface Skin(c)

Remainder

1950-

1970

(Type 

I)(b) Annual All 0.108 0.0650 0.032 (d) 0.0044 0.0650 0.0650 0.032 (d) 0.032 (d) 0.032 (d) 0.0650 0.151 0.0650

1971 -

7/1985

(Type II) Annual All 0.044 0.0268 0.00014 0.0027 0.0268 0.0268 0.00014 0.00014 0.00014 0.0268 0.062 0.0268

8/1985 -

5/199

(Type 

III)

Annual

Over 50 

years old

Biennial

40-49 

years old

Every 3rd

year

Under 40 

years old 0.033 0.0201 0.00011 0.0020 0.0201 0.0201 0.00011 0.00011 0.00011 0.0201 0.046 0.0201

1/1999 -

2003

(Type 

IV)

Biennial

Over 50 

years old

Every 3rd

year

40-49 

years old

Every 5th

year

Under 40 

years old 0.033 0.0201 0.00011 0.0020 0.0201 0.0201 0.00011 0.00011 0.00011 0.0201 0.046 0.0201 a Organ doses for lateral chests can be estimated at 2.5 times the organ doses for PA chests (Ron Kathren and Vern Shockley)

b Doses from Type I equipment are estimated at 2.5 times the organ doses from Type II equipment (Ron Kathren and Vern 

Shockley)

c Entrance skin dose is entrance skin exposure calculated from air kerma, multiplied by a backscatter factor of 1.4 from NCRP 102, 

Table B-8.

(d)No collimation assumed prior to 1970. For organs not in the chest cavity, other ICRP 34 dose conversion factors were used 

(abdomen, skull, and cervical spine).
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 146 of 188

ATTACHMENT C

OCCUPATIONAL ENVIRONMENTAL DOSE

C.0 Occupational Environmental Dose

Unmonitored workers can be exposed to occupational doses internally from the 

resuspension of radioactive materials in soil and on-site radiation releases into the air, 

and externally from ambient radiation and releases of radioactive noble gases into air.

C.1 Internal Dose from On-Site Concentrations of Radionuclides

C.1.1. On-Site Releases to Air

If a worker spent significant time in areas for which radionuclide concentrations were 

calculated and the worker was not included in the internal dosimetry program, the data in 

Tables C-01 through C-15 may be used to estimate unmonitored intakes. These tables 

show average derived intake (assuming a 2,400 cubic meters per year ventilation Rate) 

at the 50th percentile and the geometric standard deviation, based on the 95th percentile 

source term for 3

H, 131I, Pu, and U, in each SRS area by year. If the worker worked in 

multiple areas on the site, the site average annual intakes in Table C-16 should be used. 

Table C-17 shows the maximum annual intakes of these radionuclides at any SRS area 

by year. The decision to use these maximum air concentrations to estimate unmonitored 

intake should be made cautiously because the worker’s intake could be significantly 

overestimated. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 147 of 188

Table C-01 A Area Annual Intakes and geometric standard deviations, (Bq per year)

Tritium Iodine-131 Plutonium Uranium

Year 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD

1955 2.73E+03 1.46 1.01E+02 1.79 9.54E-01 1.82 9.57E-02 1.82

1956 3.81E+04 1.26 2.26E+03 1.66 3.77E-03 1.82 4.39E-02 1.82

1957 1.25E+05 1.23 4.03E+02 1.89 5.04E-03 1.82 7.24E-04 1.82

1958 1.69E+05 1.21 2.60E+01 2.14 2.48E-03 1.26 4.77E-04 1.26

1959 6.21E+04 1.23 2.93E+02 1.86 2.80E-03 1.26 2.36E-03 1.26

1960 5.48E+04 1.24 1.33E+01 2.14 8.60E-03 1.26 6.60E-03 1.26

1961 5.01E+04 1.23 1.57E+02 1.81 1.34E-03 1.26 3.28E-03 1.26

1962 6.17E+04 1.26 4.87E+00 2.86 1.05E-03 1.26 2.32E-03 1.26

1963 5.94E+04 1.31 2.49E+00 3.19 3.57E-04 1.26 8.08E-03 1.26

1964 8.64E+04 1.30 1.12E+00 1.62 5.00E-04 1.26 1.40E-02 1.26

1965 4.22E+04 1.28 1.71E+00 1.36 1.39E-03 1.26 1.39E-02 1.26

1966 3.67E+04 1.28 2.92E+00 1.20 8.58E-04 1.26 5.47E-03 1.26

1967 3.86E+04 1.27 2.12E+00 1.57 9.09E-04 1.26 5.10E-03 1.26

1968 4.54E+04 1.27 2.33E+00 1.53 5.91E-04 1.26 5.95E-03 1.26

1969 2.77E+04 1.26 3.56E+00 1.50 1.22E-01 1.26 1.62E-02 1.26

1970 2.91E+04 1.25 3.13E+00 1.32 2.44E-03 1.26 4.28E-03 1.26

1971 3.84E+04 1.24 2.16E+00 1.22 2.42E-03 1.26 1.22E-03 1.26

1972 4.62E+04 1.23 2.68E-01 1.51 1.63E-03 1.26 1.90E-03 1.26

1973 3.16E+04 1.24 1.68E-01 1.28 1.86E-03 1.26 1.27E-03 1.26

1974 4.98E+04 1.51 1.65E-01 1.16 6.57E-04 1.26 2.33E-03 1.26

1975 2.77E+04 1.39 1.06E-02 1.17 1.99E-04 1.26 1.27E-03 1.26

1976 1.52E+04 1.26 1.34E-02 1.17 1.33E-03 1.26 1.33E-03 1.26

1977 2.13E+04 1.26 5.43E-03 1.25 4.13E-04 1.26 3.95E-04 1.26

1978 1.94E+04 1.28 5.66E-03 1.18 5.73E-04 1.26 7.88E-04 1.26

1979 1.83E+04 1.26 7.65E-03 1.32 1.58E-04 1.26 6.29E-04 1.26

1980 1.67E+04 1.25 2.51E-03 1.54 3.31E-04 1.26 1.04E-03 1.26

1981 2.07E+04 1.25 4.51E-03 1.44 6.12E-04 1.26 1.51E-03 1.26

1982 2.42E+04 1.24 1.02E-02 1.44 5.60E-04 1.26 2.42E-03 1.26

1983 3.48E+04 1.25 7.77E-03 1.34 2.92E-04 1.26 1.20E-03 1.26

1984 4.42E+04 1.25 2.50E-02 1.25 1.75E-04 1.26 5.62E-04 1.26

1985 3.77E+04 1.25 5.83E-03 1.43 1.03E-04 1.26 6.52E-04 1.26

1986 2.25E+04 1.23 2.45E-03 1.34 1.88E-04 1.26 3.80E-04 1.26

1987 3.20E+04 1.32 1.75E-03 2.04 1.87E-04 1.26 2.28E-03 1.26

1988 2.03E+04 1.23 8.78E-05 2.34 1.15E-04 1.26 3.87E-04 1.26

1989 1.51E+04 1.16 3.13E-05 1.15 1.57E-04 1.26 1.35E-03 1.26

1990 1.42E+04 1.16 1.04E-03 1.99 2.27E-04 1.82 3.59E-04 1.82

1991 1.08E+04 1.36 8.14E-04 1.99 5.09E-04 1.82 1.87E-04 1.82

1992 9.35E+03 1.19 2.39E-06 1.99 5.89E-05 1.82 1.20E-04 1.82

1993 1.03E+04 2.14 3.49E-04 1.99 1.32E-04 1.82 1.46E-04 1.82

1994 8.70E+03 1.71 1.28E-06 1.99 1.34E-04 1.82 1.81E-04 1.82

1995 5.32E+03 1.51 3.39E-04 1.99 1.13E-04 1.82 1.22E-04 1.82

1996 2.89E+03 1.58 8.00E-04 1.99 4.75E-05 1.82 1.33E-04 1.82

1997 3.24E+03 1.53 1.34E-04 1.99 3.94E-06 1.82 6.42E-06 1.82

1998 3.56E+03 1.48 1.68E-06 1.99 1.15E-05 1.82 3.76E-06 1.82

1999 2.27E+02 2.47 1.68E-06 1.99 4.07E-05 1.82 3.81E-06 1.82

2000 2.17E+02 2.27 1.11E-03 1.99 4.10E-05 1.82 8.12E-06 1.82

2001 2.26E+02 1.94 1.38E-03 1.99 8.30E-06 1.82 1.07E-05 1.82
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 148 of 188

Table C-02 C Reactor Area Annual Intakes and geometric standard deviations, (Bq per

year)

Tritium Iodine-131 Plutonium Uranium

Year 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD

1955 1.23E+04 1.42 2.33E+02 1.83 2.22E+00 1.82 2.16E-01 1.82

1956 1.39E+05 1.27 5.12E+03 1.67 8.81E-03 1.82 1.03E-01 1.82

1957 3.81E+05 1.23 9.29E+02 1.98 1.17E-02 1.82 1.69E-03 1.82

1958 5.93E+05 1.22 9.30E+01 2.74 5.78E-03 1.26 1.10E-03 1.26

1959 5.11E+05 1.23 7.12E+02 1.95 6.53E-03 1.26 5.52E-03 1.26

1960 5.86E+05 1.28 3.80E+01 2.43 2.02E-02 1.26 1.55E-02 1.26

1961 3.77E+05 1.26 3.81E+02 1.88 3.11E-03 1.26 7.64E-03 1.26

1962 4.42E+05 1.27 1.83E+01 3.08 2.40E-03 1.26 5.40E-03 1.26

1963 5.90E+05 1.32 1.60E+01 3.40 8.29E-04 1.26 1.88E-02 1.26

1964 8.46E+05 1.28 3.43E+00 2.04 1.13E-03 1.26 3.24E-02 1.26

1965 7.73E+05 1.32 4.70E+00 1.70 3.09E-03 1.26 3.26E-02 1.26

1966 6.04E+05 1.28 9.26E+00 1.80 1.91E-03 1.26 1.28E-02 1.26

1967 6.39E+05 1.28 7.38E+00 2.15 2.03E-03 1.26 1.19E-02 1.26

1968 8.05E+05 1.29 7.21E+00 2.05 1.33E-03 1.26 1.34E-02 1.26

1969 4.86E+05 1.30 9.99E+00 1.84 2.66E-01 1.26 3.69E-02 1.26

1970 4.53E+05 1.29 1.16E+01 2.14 5.39E-03 1.26 9.95E-03 1.26

1971 5.28E+05 1.27 7.00E+00 1.90 5.35E-03 1.26 2.78E-03 1.26

1972 5.62E+05 1.26 1.80E+00 2.72 3.59E-03 1.26 4.38E-03 1.26

1973 4.36E+05 1.27 4.88E-01 1.74 4.08E-03 1.26 2.92E-03 1.26

1974 5.08E+05 1.35 4.07E-01 1.31 1.47E-03 1.26 5.47E-03 1.26

1975 2.79E+05 1.31 2.50E-02 1.19 4.37E-04 1.26 2.98E-03 1.26

1976 2.58E+05 1.26 3.47E-02 1.41 2.94E-03 1.26 3.13E-03 1.26

1977 4.96E+05 1.29 2.10E-02 2.09 9.08E-04 1.26 9.17E-04 1.26

1978 3.95E+05 1.29 1.39E-02 1.30 1.26E-03 1.26 1.83E-03 1.26

1979 3.55E+05 1.33 2.05E-02 1.65 3.57E-04 1.26 1.46E-03 1.26

1980 3.80E+05 1.28 8.03E-03 2.05 7.33E-04 1.26 2.42E-03 1.26

1981 2.51E+05 1.27 2.11E-02 2.41 1.36E-03 1.26 3.48E-03 1.26

1982 3.88E+05 1.31 3.39E-02 2.04 1.29E-03 1.26 5.65E-03 1.26

1983 3.88E+05 1.29 2.64E-02 2.01 6.57E-04 1.26 2.80E-03 1.26

1984 7.16E+05 1.30 5.95E-02 1.29 4.01E-04 1.26 1.31E-03 1.26

1985 9.53E+05 1.28 1.38E-02 1.44 2.39E-04 1.26 1.52E-03 1.26

1986 2.38E+05 1.27 6.03E-03 1.45 4.16E-04 1.26 8.85E-04 1.26

1987 8.64E+04 1.30 4.22E-03 2.07 4.15E-04 1.26 5.35E-03 1.26

1988 6.31E+04 1.24 2.50E-04 2.50 2.66E-04 1.26 9.05E-04 1.26

1989 4.67E+04 1.13 7.34E-05 1.15 3.64E-04 1.26 3.17E-03 1.26

1990 4.22E+04 1.12 2.43E-03 1.99 5.16E-04 1.82 8.18E-04 1.82

1991 3.22E+04 1.26 1.91E-03 1.99 1.16E-03 1.82 4.25E-04 1.82

1992 2.69E+04 1.15 5.23E-06 1.99 1.34E-04 1.82 2.81E-04 1.82

1993 2.75E+04 2.73 7.95E-04 1.99 2.99E-04 1.82 3.42E-04 1.82

1994 2.10E+04 2.03 2.79E-06 1.99 2.99E-04 1.82 4.24E-04 1.82

1995 1.77E+04 1.58 7.80E-04 1.99 2.61E-04 1.82 2.87E-04 1.82

1996 9.34E+03 1.69 1.86E-03 1.99 1.08E-04 1.82 3.11E-04 1.82

1997 1.10E+04 1.60 2.92E-04 1.99 8.87E-06 1.82 1.44E-05 1.82

1998 9.77E+03 1.55 3.95E-06 1.99 2.59E-05 1.82 8.80E-06 1.82

1999 2.25E+03 2.00 3.95E-06 1.99 9.04E-05 1.82 8.88E-06 1.82

2000 4.99E+03 1.62 2.46E-03 1.99 9.14E-05 1.82 1.90E-05 1.82

2001 5.19E+03 1.52 3.06E-03 1.99 1.83E-05 1.82 2.51E-05 1.82
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 149 of 188

Table C-03 D Area Annual Intakes and geometric standard deviations, (Bq per year)

Tritium Iodine-131 Plutonium Uranium

Year 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD

1955 2.12E+05 1.18 7.48E+01 1.80 6.96E-01 1.82 7.21E-02 1.82

1956 2.36E+05 1.08 1.70E+03 1.67 2.74E-03 1.82 3.18E-02 1.82

1957 3.09E+05 1.09 3.16E+02 1.91 3.66E-03 1.82 5.27E-04 1.82

1958 3.35E+05 1.11 2.29E+01 2.33 1.80E-03 1.26 3.53E-04 1.26

1959 5.40E+05 1.04 2.16E+02 1.88 2.04E-03 1.26 1.71E-03 1.26

1960 3.57E+05 1.13 1.07E+01 2.30 6.22E-03 1.26 4.77E-03 1.26

1961 2.92E+05 1.13 1.15E+02 1.84 9.84E-04 1.26 2.38E-03 1.26

1962 2.56E+05 1.17 6.02E+00 3.10 7.81E-04 1.26 1.69E-03 1.26

1963 2.60E+05 1.22 2.74E+00 3.30 2.61E-04 1.26 5.90E-03 1.26

1964 2.92E+05 1.22 1.06E+00 2.05 3.76E-04 1.26 1.03E-02 1.26

1965 2.58E+05 1.21 1.40E+00 1.64 1.06E-03 1.26 1.01E-02 1.26

1966 4.68E+05 1.13 2.22E+00 1.32 6.54E-04 1.26 3.97E-03 1.26

1967 4.70E+05 1.11 2.16E+00 2.09 6.92E-04 1.26 3.70E-03 1.26

1968 2.46E+05 1.17 2.26E+00 2.08 4.46E-04 1.26 4.49E-03 1.26

1969 2.98E+05 1.13 3.48E+00 2.02 9.50E-02 1.26 1.20E-02 1.26

1970 2.36E+05 1.12 2.54E+00 1.58 1.88E-03 1.26 3.13E-03 1.26

1971 2.31E+05 1.18 1.61E+00 1.31 1.87E-03 1.26 9.07E-04 1.26

1972 2.25E+05 1.18 2.29E-01 1.84 1.27E-03 1.26 1.40E-03 1.26

1973 1.97E+05 1.19 1.33E-01 1.51 1.45E-03 1.26 9.33E-04 1.26

1974 1.64E+05 1.31 1.20E-01 1.18 5.01E-04 1.26 1.69E-03 1.26

1975 1.72E+05 1.21 7.84E-03 1.23 1.54E-04 1.26 9.20E-04 1.26

1976 1.02E+05 1.20 9.76E-03 1.19 1.03E-03 1.26 9.65E-04 1.26

1977 1.06E+05 1.21 4.12E-03 1.37 3.19E-04 1.26 2.89E-04 1.26

1978 1.05E+05 1.20 4.19E-03 1.24 4.44E-04 1.26 5.75E-04 1.26

1979 8.35E+04 1.26 5.84E-03 1.47 1.19E-04 1.26 4.59E-04 1.26

1980 7.89E+04 1.21 2.11E-03 1.82 2.54E-04 1.26 7.58E-04 1.26

1981 8.01E+04 1.21 3.65E-03 1.68 4.70E-04 1.26 1.11E-03 1.26

1982 8.46E+04 1.23 8.26E-03 1.68 4.13E-04 1.26 1.76E-03 1.26

1983 9.64E+04 1.22 6.03E-03 1.50 2.21E-04 1.26 8.67E-04 1.26

1984 1.04E+05 1.23 1.87E-02 1.34 1.29E-04 1.26 4.10E-04 1.26

1985 1.01E+05 1.22 4.70E-03 1.64 7.59E-05 1.26 4.76E-04 1.26

1986 5.70E+04 1.21 2.08E-03 1.66 1.44E-04 1.26 2.77E-04 1.26

1987 3.76E+04 1.29 1.79E-03 2.38 1.43E-04 1.26 1.65E-03 1.26

1988 3.11E+04 1.24 1.20E-04 2.76 8.45E-05 1.26 2.81E-04 1.26

1989 2.59E+04 1.12 2.26E-05 1.15 1.14E-04 1.26 9.78E-04 1.26

1990 2.55E+04 1.11 7.50E-04 1.99 1.69E-04 1.82 2.68E-04 1.82

1991 2.32E+04 1.16 5.88E-04 1.99 3.80E-04 1.82 1.39E-04 1.82

1992 1.08E+04 1.17 1.87E-06 1.99 4.41E-05 1.82 8.71E-05 1.82

1993 2.16E+04 2.37 2.61E-04 1.99 9.84E-05 1.82 1.06E-04 1.82

1994 1.49E+04 1.84 9.97E-07 1.99 1.02E-04 1.82 1.32E-04 1.82

1995 1.21E+04 1.35 2.51E-04 1.99 8.33E-05 1.82 8.84E-05 1.82

1996 1.01E+04 1.34 5.83E-04 1.99 3.56E-05 1.82 9.59E-05 1.82

1997 1.09E+04 1.28 1.04E-04 1.99 2.98E-06 1.82 4.88E-06 1.82

1998 1.23E+04 1.22 1.22E-06 1.99 8.72E-06 1.82 2.72E-06 1.82

1999 5.82E+03 1.26 1.22E-06 1.99 3.12E-05 1.82 2.77E-06 1.82

2000 5.80E+03 1.21 8.55E-04 1.99 3.13E-05 1.82 5.90E-06 1.82

2001 5.78E+03 1.16 1.05E-03 1.99 6.42E-06 1.82 7.72E-06 1.82
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 150 of 188

Table C-04 E Area Annual Intakes and geometric standard deviations, (Bq per year)

Tritium Iodine-131 Plutonium Uranium

Year 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD

1955 1.64E+04 1.50 7.40E+02 1.79 6.86E+00 1.82 7.20E-01 1.82

1956 2.61E+05 1.27 1.70E+04 1.66 2.70E-02 1.82 3.13E-01 1.82

1957 9.25E+05 1.23 3.14E+03 1.88 3.61E-02 1.82 5.19E-03 1.82

1958 1.25E+06 1.21 1.93E+02 2.03 1.77E-02 1.26 3.50E-03 1.26

1959 4.36E+05 1.23 2.10E+03 1.84 2.01E-02 1.26 1.68E-02 1.26

1960 3.54E+05 1.25 8.96E+01 2.05 6.11E-02 1.26 4.69E-02 1.26

1961 3.43E+05 1.23 1.11E+03 1.80 9.72E-03 1.26 2.35E-02 1.26

1962 4.02E+05 1.27 1.85E+01 2.28 7.76E-03 1.26 1.66E-02 1.26

1963 3.97E+05 1.27 8.40E+00 2.80 2.57E-03 1.26 5.83E-02 1.26

1964 5.27E+05 1.32 6.99E+00 1.33 3.75E-03 1.26 1.02E-01 1.26

1965 2.21E+05 1.29 1.16E+01 1.24 1.07E-02 1.26 9.89E-02 1.26

1966 2.07E+05 1.29 2.04E+01 1.15 6.56E-03 1.26 3.90E-02 1.26

1967 2.06E+05 1.27 1.33E+01 1.30 6.94E-03 1.26 3.64E-02 1.26

1968 2.68E+05 1.27 1.51E+01 1.30 4.46E-03 1.26 4.49E-02 1.26

1969 1.71E+05 1.24 2.31E+01 1.27 9.61E-01 1.26 1.20E-01 1.26

1970 1.76E+05 1.22 2.15E+01 1.22 1.89E-02 1.26 3.08E-02 1.26

1971 2.39E+05 1.22 1.51E+01 1.19 1.88E-02 1.26 9.02E-03 1.26

1972 3.08E+05 1.22 1.78E+00 1.34 1.28E-02 1.26 1.39E-02 1.26

1973 2.02E+05 1.22 1.16E+00 1.20 1.46E-02 1.26 9.24E-03 1.26

1974 3.44E+05 1.54 1.17E+00 1.16 5.02E-03 1.26 1.66E-02 1.26

1975 1.81E+05 1.40 7.50E-02 1.16 1.56E-03 1.26 9.04E-03 1.26

1976 9.28E+04 1.23 9.48E-02 1.16 1.04E-02 1.26 9.48E-03 1.26

1977 1.30E+05 1.24 3.78E-02 1.20 3.22E-03 1.26 2.86E-03 1.26

1978 1.21E+05 1.27 4.00E-02 1.16 4.49E-03 1.26 5.67E-03 1.26

1979 1.11E+05 1.23 5.24E-02 1.21 1.19E-03 1.26 4.53E-03 1.26

1980 9.95E+04 1.22 1.62E-02 1.30 2.56E-03 1.26 7.47E-03 1.26

1981 1.41E+05 1.23 3.00E-02 1.28 4.73E-03 1.26 1.10E-02 1.26

1982 1.58E+05 1.22 6.73E-02 1.27 4.09E-03 1.26 1.73E-02 1.26

1983 2.37E+05 1.23 5.29E-02 1.23 2.22E-03 1.26 8.52E-03 1.26

1984 2.99E+05 1.23 1.74E-01 1.19 1.28E-03 1.26 4.05E-03 1.26

1985 2.37E+05 1.23 3.85E-02 1.25 7.51E-04 1.26 4.69E-03 1.26

1986 1.59E+05 1.21 1.65E-02 1.22 1.45E-03 1.26 2.74E-03 1.26

1987 2.40E+05 1.31 9.16E-03 1.57 1.44E-03 1.26 1.62E-02 1.26

1988 1.40E+05 1.22 4.07E-04 1.81 8.36E-04 1.26 2.77E-03 1.26

1989 9.42E+04 1.20 2.22E-04 1.15 1.13E-03 1.26 9.62E-03 1.26

1990 9.17E+04 1.20 7.37E-03 1.99 1.68E-03 1.82 2.67E-03 1.82

1991 6.96E+04 1.44 5.78E-03 1.99 3.78E-03 1.82 1.39E-03 1.82

1992 6.38E+04 1.21 1.89E-05 1.99 4.40E-04 1.82 8.56E-04 1.82

1993 7.31E+04 1.66 2.59E-03 1.99 9.81E-04 1.82 1.04E-03 1.82

1994 6.42E+04 1.51 1.01E-05 1.99 1.02E-03 1.82 1.30E-03 1.82

1995 3.96E+04 1.44 2.49E-03 1.99 8.25E-04 1.82 8.68E-04 1.82

1996 2.08E+04 1.47 5.75E-03 1.99 3.54E-04 1.82 9.42E-04 1.82

1997 2.44E+04 1.44 1.05E-03 1.99 2.98E-05 1.82 4.89E-05 1.82

1998 2.72E+04 1.43 1.20E-05 1.99 8.73E-05 1.82 2.68E-05 1.82

1999 1.27E+03 1.92 1.20E-05 1.99 3.14E-04 1.82 2.73E-05 1.82

2000 1.06E+03 1.89 8.61E-03 1.99 3.14E-04 1.82 5.80E-05 1.82

2001 1.22E+03 1.68 1.06E-02 1.99 6.48E-05 1.82 7.59E-05 1.82
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 151 of 188

Table C-05 F Area Annual Intakes and geometric standard deviations, (Bq per year)

Tritium Iodine-131 Plutonium Uranium

Year 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD

1955 4.50E+04 1.61 2.06E+03 1.79 2.46E+01 1.82 1.46E+00 1.82

1956 1.02E+06 1.26 3.52E+04 1.66 1.02E-01 1.82 1.22E+00 1.82

1957 1.80E+06 1.21 2.11E+03 1.88 1.35E-01 1.82 1.94E-02 1.82

1958 2.37E+06 1.20 9.54E+01 2.09 6.74E-02 1.26 1.01E-02 1.26

1959 2.23E+05 1.24 7.58E+03 1.84 7.48E-02 1.26 6.73E-02 1.26

1960 1.88E+05 1.27 3.57E+02 2.00 2.47E-01 1.26 1.89E-01 1.26

1961 1.75E+05 1.25 4.50E+03 1.80 3.26E-02 1.26 8.75E-02 1.26

1962 2.06E+05 1.29 5.07E+01 1.56 2.06E-02 1.26 6.20E-02 1.26

1963 2.05E+05 1.32 1.73E+01 1.95 9.06E-03 1.26 2.05E-01 1.26

1964 2.78E+05 1.35 2.69E+01 1.18 7.87E-03 1.26 3.39E-01 1.26

1965 1.34E+05 1.33 4.62E+01 1.18 1.41E-02 1.26 3.99E-01 1.26

1966 1.19E+05 1.32 8.21E+01 1.12 9.71E-03 1.26 1.53E-01 1.26

1967 1.23E+05 1.31 5.16E+01 1.16 1.08E-02 1.26 1.43E-01 1.26

1968 1.55E+05 1.30 5.91E+01 1.19 8.64E-03 1.26 8.70E-02 1.26

1969 9.95E+04 1.26 9.07E+01 1.18 4.65E-01 1.26 3.07E-01 1.26

1970 1.02E+05 1.24 8.60E+01 1.17 1.98E-02 1.26 1.10E-01 1.26

1971 1.36E+05 1.24 6.09E+01 1.16 1.84E-02 1.26 2.31E-02 1.26

1972 1.67E+05 1.23 6.88E+00 1.21 9.56E-03 1.26 4.24E-02 1.26

1973 1.13E+05 1.23 4.63E+00 1.17 8.88E-03 1.26 2.91E-02 1.26

1974 1.82E+05 1.52 4.72E+00 1.15 7.66E-03 1.26 6.73E-02 1.26

1975 9.80E+04 1.39 2.99E-01 1.16 1.10E-03 1.26 3.67E-02 1.26

1976 5.37E+04 1.25 3.85E-01 1.16 9.30E-03 1.26 3.81E-02 1.26

1977 7.89E+04 1.25 1.50E-01 1.17 2.60E-03 1.26 9.85E-03 1.26

1978 7.12E+04 1.28 1.62E-01 1.16 3.22E-03 1.26 2.02E-02 1.26

1979 6.58E+04 1.25 2.08E-01 1.17 2.41E-03 1.26 1.63E-02 1.26

1980 6.06E+04 1.24 6.24E-02 1.19 2.91E-03 1.26 2.71E-02 1.26

1981 7.55E+04 1.24 1.18E-01 1.19 5.60E-03 1.26 3.49E-02 1.26

1982 8.86E+04 1.24 2.65E-01 1.18 1.27E-02 1.26 6.68E-02 1.26

1983 1.27E+05 1.24 2.11E-01 1.17 3.98E-03 1.26 3.35E-02 1.26

1984 1.65E+05 1.24 7.01E-01 1.16 3.63E-03 1.26 1.43E-02 1.26

1985 1.43E+05 1.25 1.52E-01 1.17 2.43E-03 1.26 1.69E-02 1.26

1986 8.26E+04 1.23 6.60E-02 1.17 1.65E-03 1.26 9.88E-03 1.26

1987 1.16E+05 1.32 3.26E-02 1.26 1.80E-03 1.26 6.55E-02 1.26

1988 7.04E+04 1.23 1.33E-03 1.36 2.72E-03 1.26 1.07E-02 1.26

1989 4.88E+04 1.20 9.05E-04 1.15 3.96E-03 1.26 3.84E-02 1.26

1990 4.70E+04 1.19 3.00E-02 1.99 4.15E-03 1.82 6.58E-03 1.82

1991 3.58E+04 1.41 2.36E-02 1.99 9.31E-03 1.82 3.42E-03 1.82

1992 3.23E+04 1.21 9.02E-06 1.99 1.01E-03 1.82 3.37E-03 1.82

1993 3.66E+04 1.90 6.40E-03 1.99 2.31E-03 1.82 4.09E-03 1.82

1994 3.15E+04 1.61 4.82E-06 1.99 1.62E-03 1.82 5.00E-03 1.82

1995 1.94E+04 1.47 7.31E-03 1.99 2.58E-03 1.82 3.52E-03 1.82

1996 1.03E+04 1.53 2.08E-02 1.99 8.42E-04 1.82 3.84E-03 1.82

1997 1.19E+04 1.49 5.23E-04 1.99 5.58E-05 1.82 7.91E-05 1.82

1998 1.32E+04 1.46 4.88E-05 1.99 1.55E-04 1.82 1.06E-04 1.82

1999 7.18E+02 2.25 4.88E-05 1.99 3.77E-04 1.82 1.02E-04 1.82

2000 6.84E+02 2.12 9.26E-03 1.99 4.46E-04 1.82 2.23E-04 1.82

2001 7.49E+02 1.84 1.41E-02 1.99 5.22E-05 1.82 3.08E-04 1.82
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 152 of 188

Table C-06 H Area Annual Intakes and geometric standard deviations, (Bq per year)

Tritium Iodine-131 Plutonium Uranium

Year 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD

1955 3.04E+04 1.28 1.51E+03 1.79 7.08E+00 1.82 2.17E+00 1.82

1956 1.46E+05 1.25 5.03E+04 1.66 2.05E-02 1.82 1.92E-01 1.82

1957 2.88E+06 1.24 1.50E+04 1.87 2.94E-02 1.82 4.23E-03 1.82

1958 3.93E+06 1.20 9.41E+02 1.98 1.30E-02 1.26 6.67E-03 1.26

1959 2.12E+06 1.22 2.05E+03 1.84 1.70E-02 1.26 7.84E-03 1.26

1960 1.68E+06 1.22 3.74E+01 2.13 2.60E-02 1.26 2.09E-02 1.26

1961 1.72E+06 1.20 4.46E+02 1.81 1.29E-02 1.26 1.97E-02 1.26

1962 2.01E+06 1.24 1.29E+01 2.81 1.73E-02 1.26 1.38E-02 1.26

1963 1.97E+06 1.22 7.78E+00 3.22 2.83E-03 1.26 6.41E-02 1.26

1964 2.54E+06 1.28 3.05E+00 1.58 1.10E-02 1.26 1.36E-01 1.26

1965 1.00E+06 1.22 4.76E+00 1.33 4.19E-02 1.26 4.35E-02 1.26

1966 1.01E+06 1.23 8.37E+00 1.19 2.45E-02 1.26 2.34E-02 1.26

1967 9.22E+05 1.22 5.71E+00 1.50 2.52E-02 1.26 2.04E-02 1.26

1968 1.27E+06 1.22 6.43E+00 1.46 1.40E-02 1.26 1.41E-01 1.26

1969 8.03E+05 1.21 9.74E+00 1.43 4.82E+00 1.26 2.80E-01 1.26

1970 8.58E+05 1.19 8.72E+00 1.29 8.10E-02 1.26 3.25E-02 1.26

1971 1.10E+06 1.21 6.23E+00 1.21 8.23E-02 1.26 2.12E-02 1.26

1972 1.47E+06 1.20 7.70E-01 1.46 5.97E-02 1.26 2.37E-02 1.26

1973 1.00E+06 1.19 4.93E-01 1.25 7.10E-02 1.26 1.46E-02 1.26

1974 1.68E+06 1.53 5.02E-01 1.16 1.84E-02 1.26 6.87E-03 1.26

1975 8.69E+05 1.39 3.69E-02 1.17 7.37E-03 1.26 3.73E-03 1.26

1976 4.69E+05 1.20 3.81E-02 1.17 4.66E-02 1.26 4.40E-03 1.26

1977 5.77E+05 1.21 1.75E-02 1.23 1.48E-02 1.26 3.41E-03 1.26

1978 5.58E+05 1.24 1.67E-02 1.17 2.11E-02 1.26 5.96E-03 1.26

1979 5.06E+05 1.20 2.40E-02 1.29 3.61E-03 1.26 4.54E-03 1.26

1980 4.50E+05 1.19 7.95E-03 1.47 1.07E-02 1.26 7.20E-03 1.26

1981 7.13E+05 1.21 1.27E-02 1.41 1.94E-02 1.26 1.70E-02 1.26

1982 7.41E+05 1.20 2.88E-02 1.41 6.65E-03 1.26 1.14E-02 1.26

1983 1.16E+06 1.22 2.22E-02 1.32 7.37E-03 1.26 4.88E-03 1.26

1984 1.41E+06 1.21 7.20E-02 1.24 2.57E-03 1.26 4.42E-03 1.26

1985 1.05E+06 1.21 1.68E-02 1.39 1.11E-03 1.26 4.70E-03 1.26

1986 8.77E+05 1.19 6.98E-03 1.30 6.06E-03 1.26 2.70E-03 1.26

1987 1.28E+06 1.30 4.76E-03 1.98 5.80E-03 1.26 7.14E-03 1.26

1988 7.17E+05 1.21 2.28E-04 2.27 1.20E-03 1.26 1.86E-03 1.26

1989 4.38E+05 1.21 8.65E-05 1.15 1.27E-03 1.26 4.80E-03 1.26

1990 4.33E+05 1.20 2.86E-03 1.99 4.15E-03 1.82 6.58E-03 1.82

1991 3.28E+05 1.46 2.24E-03 1.99 9.31E-03 1.82 3.42E-03 1.82

1992 3.06E+05 1.21 9.47E-05 1.99 1.18E-03 1.82 4.92E-04 1.82

1993 3.54E+05 1.45 6.38E-03 1.99 2.56E-03 1.82 6.06E-04 1.82

1994 3.16E+05 1.42 5.06E-05 1.99 3.66E-03 1.82 8.59E-04 1.82

1995 1.95E+05 1.41 4.61E-03 1.99 1.33E-03 1.82 3.70E-04 1.82

1996 1.01E+05 1.42 5.67E-03 1.99 9.11E-04 1.82 3.71E-04 1.82

1997 1.21E+05 1.41 5.27E-03 1.99 9.60E-05 1.82 1.74E-04 1.82

1998 1.35E+05 1.41 4.64E-06 1.99 2.92E-04 1.82 1.39E-05 1.82

1999 5.55E+03 1.52 4.64E-06 1.99 1.29E-03 1.82 2.29E-05 1.82

2000 4.08E+03 1.53 3.66E-02 1.99 1.20E-03 1.82 4.05E-05 1.82

2001 4.98E+03 1.47 4.14E-02 1.99 2.98E-04 1.82 3.19E-05 1.82
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 153 of 188

Table C-07 S Area Annual Intakes and geometric standard deviations, (Bq per year)

Tritium Iodine-131 Plutonium Uranium

Year 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD

1955 1.31E+04 1.35 6.23E+02 1.79 3.68E+00 1.82 8.18E-01 1.82

1956 1.03E+05 1.27 1.90E+04 1.66 1.22E-02 1.82 1.28E-01 1.82

1957 1.08E+06 1.24 5.24E+03 1.87 1.70E-02 1.82 2.44E-03 1.82

1958 1.47E+06 1.21 3.30E+02 2.00 7.90E-03 1.26 2.80E-03 1.26

1959 7.35E+05 1.23 1.09E+03 1.85 9.65E-03 1.26 6.11E-03 1.26

1960 5.83E+05 1.23 3.23E+01 2.13 2.15E-02 1.26 1.68E-02 1.26

1961 5.82E+05 1.21 3.83E+02 1.81 6.06E-03 1.26 1.12E-02 1.26

1962 6.79E+05 1.26 1.12E+01 2.81 6.96E-03 1.26 7.88E-03 1.26

1963 6.68E+05 1.24 6.87E+00 3.23 1.43E-03 1.26 3.24E-02 1.26

1964 8.72E+05 1.30 2.64E+00 1.56 4.17E-03 1.26 6.37E-02 1.26

1965 3.37E+05 1.25 4.10E+00 1.32 1.51E-02 1.26 3.52E-02 1.26

1966 3.28E+05 1.26 7.14E+00 1.18 8.88E-03 1.26 1.58E-02 1.26

1967 3.16E+05 1.24 4.94E+00 1.48 9.19E-03 1.26 1.43E-02 1.26

1968 4.22E+05 1.25 5.51E+00 1.45 5.24E-03 1.26 5.28E-02 1.26

1969 2.67E+05 1.23 8.40E+00 1.42 1.67E+00 1.26 1.12E-01 1.26

1970 2.78E+05 1.20 7.53E+00 1.28 2.87E-02 1.26 1.67E-02 1.26

1971 3.78E+05 1.21 5.28E+00 1.21 2.91E-02 1.26 8.42E-03 1.26

1972 5.01E+05 1.21 6.45E-01 1.45 2.09E-02 1.26 1.03E-02 1.26

1973 3.28E+05 1.20 4.12E-01 1.25 2.48E-02 1.26 6.48E-03 1.26

1974 5.71E+05 1.55 4.13E-01 1.16 6.70E-03 1.26 5.78E-03 1.26

1975 2.97E+05 1.40 2.80E-02 1.17 2.58E-03 1.26 3.14E-03 1.26

1976 1.48E+05 1.22 3.27E-02 1.17 1.64E-02 1.26 3.45E-03 1.26

1977 1.97E+05 1.22 1.38E-02 1.23 5.20E-03 1.26 1.67E-03 1.26

1978 1.88E+05 1.26 1.40E-02 1.17 7.40E-03 1.26 3.07E-03 1.26

1979 1.70E+05 1.21 1.93E-02 1.30 1.36E-03 1.26 2.38E-03 1.26

1980 1.51E+05 1.20 6.32E-03 1.49 3.80E-03 1.26 3.85E-03 1.26

1981 2.33E+05 1.23 1.09E-02 1.40 6.94E-03 1.26 7.60E-03 1.26

1982 2.52E+05 1.21 2.46E-02 1.41 2.93E-03 1.26 7.32E-03 1.26

1983 3.91E+05 1.22 1.89E-02 1.32 2.73E-03 1.26 3.38E-03 1.26

1984 4.82E+05 1.22 6.13E-02 1.24 1.07E-03 1.26 2.24E-03 1.26

1985 3.59E+05 1.22 1.42E-02 1.40 5.04E-04 1.26 2.47E-03 1.26

1986 2.72E+05 1.20 5.93E-03 1.30 2.16E-03 1.26 1.43E-03 1.26

1987 4.20E+05 1.31 4.10E-03 1.98 2.08E-03 1.26 5.78E-03 1.26

1988 2.40E+05 1.22 1.93E-04 2.25 5.51E-04 1.26 1.18E-03 1.26

1989 1.57E+05 1.20 7.57E-05 1.15 6.38E-04 1.26 3.60E-03 1.26

1990 1.54E+05 1.20 2.51E-03 1.99 1.63E-03 1.82 2.59E-03 1.82

1991 1.16E+05 1.45 1.97E-03 1.99 3.66E-03 1.82 1.34E-03 1.82

1992 1.08E+05 1.21 3.29E-05 1.99 4.55E-04 1.82 3.40E-04 1.82

1993 1.24E+05 1.52 2.51E-03 1.99 9.92E-04 1.82 4.16E-04 1.82

1994 1.10E+05 1.45 1.76E-05 1.99 1.33E-03 1.82 5.49E-04 1.82

1995 6.81E+04 1.42 1.95E-03 1.99 5.88E-04 1.82 3.06E-04 1.82

1996 3.56E+04 1.43 3.00E-03 1.99 3.55E-04 1.82 3.22E-04 1.82

1997 4.21E+04 1.42 1.83E-03 1.99 3.57E-05 1.82 6.35E-05 1.82

1998 4.70E+04 1.41 4.07E-06 1.99 1.08E-04 1.82 1.02E-05 1.82

1999 2.00E+03 1.67 4.07E-06 1.99 4.59E-04 1.82 1.30E-05 1.82

2000 1.52E+03 1.68 1.30E-02 1.99 4.32E-04 1.82 2.52E-05 1.82

2001 1.82E+03 1.55 1.49E-02 1.99 1.05E-04 1.82 2.66E-05 1.82
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 154 of 188

Table C-08 Z Area Annual Intakes and geometric standard deviations, (Bq per year)

Tritium Iodine-131 Plutonium Uranium

Year 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD

1955 8.30E+03 1.40 3.78E+02 1.79 2.50E+00 1.82 4.69E-01 1.82

1956 7.88E+04 1.27 1.09E+04 1.66 8.79E-03 1.82 9.53E-02 1.82

1957 6.14E+05 1.24 2.84E+03 1.88 1.20E-02 1.82 1.73E-03 1.82

1958 8.36E+05 1.21 1.80E+02 2.02 5.71E-03 1.26 1.72E-03 1.26

1959 4.00E+05 1.24 7.51E+02 1.85 6.80E-03 1.26 4.75E-03 1.26

1960 3.20E+05 1.25 2.62E+01 2.16 1.69E-02 1.26 1.31E-02 1.26

1961 3.15E+05 1.23 3.05E+02 1.82 3.95E-03 1.26 7.92E-03 1.26

1962 3.69E+05 1.27 1.02E+01 2.90 4.17E-03 1.26 5.57E-03 1.26

1963 3.62E+05 1.27 6.86E+00 3.29 9.64E-04 1.26 2.18E-02 1.26

1964 4.77E+05 1.32 2.17E+00 1.62 2.40E-03 1.26 4.15E-02 1.26

1965 1.88E+05 1.27 3.30E+00 1.35 8.35E-03 1.26 2.76E-02 1.26

1966 1.79E+05 1.28 5.68E+00 1.19 4.95E-03 1.26 1.18E-02 1.26

1967 1.77E+05 1.27 4.02E+00 1.53 5.14E-03 1.26 1.08E-02 1.26

1968 2.31E+05 1.26 4.44E+00 1.49 2.98E-03 1.26 3.01E-02 1.26

1969 1.47E+05 1.23 6.79E+00 1.46 9.03E-01 1.26 6.63E-02 1.26

1970 1.52E+05 1.21 6.02E+00 1.29 1.58E-02 1.26 1.13E-02 1.26

1971 2.10E+05 1.22 4.19E+00 1.22 1.59E-02 1.26 5.00E-03 1.26

1972 2.75E+05 1.21 5.16E-01 1.47 1.14E-02 1.26 6.42E-03 1.26

1973 1.79E+05 1.21 3.27E-01 1.27 1.34E-02 1.26 4.10E-03 1.26

1974 3.12E+05 1.55 3.25E-01 1.16 3.74E-03 1.26 4.57E-03 1.26

1975 1.64E+05 1.41 2.16E-02 1.17 1.40E-03 1.26 2.49E-03 1.26

1976 8.05E+04 1.23 2.60E-02 1.17 8.97E-03 1.26 2.68E-03 1.26

1977 1.10E+05 1.23 1.08E-02 1.24 2.83E-03 1.26 1.11E-03 1.26

1978 1.04E+05 1.26 1.11E-02 1.18 4.02E-03 1.26 2.08E-03 1.26

1979 9.49E+04 1.22 1.53E-02 1.33 7.78E-04 1.26 1.63E-03 1.26

1980 8.40E+04 1.21 5.06E-03 1.55 2.09E-03 1.26 2.65E-03 1.26

1981 1.26E+05 1.23 8.75E-03 1.43 3.82E-03 1.26 4.83E-03 1.26

1982 1.38E+05 1.21 1.99E-02 1.45 1.84E-03 1.26 5.38E-03 1.26

1983 2.13E+05 1.23 1.52E-02 1.34 1.54E-03 1.26 2.54E-03 1.26

1984 2.64E+05 1.22 4.89E-02 1.26 6.49E-04 1.26 1.51E-03 1.26

1985 2.00E+05 1.23 1.15E-02 1.44 3.22E-04 1.26 1.69E-03 1.26

1986 1.44E+05 1.21 4.75E-03 1.33 1.19E-03 1.26 9.80E-04 1.26

1987 2.24E+05 1.31 3.51E-03 2.08 1.15E-03 1.26 4.54E-03 1.26

1988 1.30E+05 1.22 1.68E-04 2.33 3.54E-04 1.26 8.66E-04 1.26

1989 8.81E+04 1.20 6.04E-05 1.15 4.28E-04 1.26 2.77E-03 1.26

1990 8.55E+04 1.20 2.00E-03 1.99 9.62E-04 1.82 1.53E-03 1.82

1991 6.44E+04 1.44 1.57E-03 1.99 2.16E-03 1.82 7.92E-04 1.82

1992 5.91E+04 1.21 1.77E-05 1.99 2.65E-04 1.82 2.56E-04 1.82

1993 6.80E+04 1.60 1.48E-03 1.99 5.80E-04 1.82 3.12E-04 1.82

1994 6.01E+04 1.48 9.48E-06 1.99 7.48E-04 1.82 4.04E-04 1.82

1995 3.71E+04 1.43 1.20E-03 1.99 3.70E-04 1.82 2.41E-04 1.82

1996 1.94E+04 1.45 2.06E-03 1.99 2.08E-04 1.82 2.57E-04 1.82

1997 2.28E+04 1.43 9.88E-04 1.99 2.03E-05 1.82 3.56E-05 1.82

1998 2.55E+04 1.42 3.25E-06 1.99 6.09E-05 1.82 7.78E-06 1.82

1999 1.13E+03 1.82 3.25E-06 1.99 2.53E-04 1.82 9.20E-06 1.82

2000 8.77E+02 1.83 7.14E-03 1.99 2.40E-04 1.82 1.84E-05 1.82

2001 1.03E+03 1.64 8.26E-03 1.99 5.70E-05 1.82 2.10E-05 1.82
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 155 of 188

Table C-09 Forest Service Headquarters (a) Calculated Air Concentration (Bq per m3) 

Tritium Iodine-131 Plutonium Uranium

Year 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD

1955 2.36E+03 1.49 9.55E+01 1.80 8.19E-01 1.82 9.94E-02 1.82

1956 3.09E+04 1.28 2.34E+03 1.66 3.15E-03 1.82 3.61E-02 1.82

1957 1.30E+05 1.24 4.88E+02 1.88 4.23E-03 1.82 6.08E-04 1.82

1958 1.75E+05 1.22 3.16E+01 2.11 2.06E-03 1.26 4.47E-04 1.26

1959 7.08E+04 1.26 2.51E+02 1.86 2.36E-03 1.26 1.91E-03 1.26

1960 5.95E+04 1.29 1.10E+01 2.18 6.95E-03 1.26 5.34E-03 1.26

1961 5.63E+04 1.27 1.27E+02 1.82 1.19E-03 1.26 2.76E-03 1.26

1962 6.70E+04 1.33 5.06E+00 2.99 1.01E-03 1.26 1.95E-03 1.26

1963 6.45E+04 1.36 2.99E+00 3.30 3.09E-04 1.26 6.99E-03 1.26

1964 8.84E+04 1.40 9.45E-01 1.71 5.15E-04 1.26 1.24E-02 1.26

1965 4.06E+04 1.36 1.39E+00 1.39 1.56E-03 1.26 1.13E-02 1.26

1966 3.60E+04 1.36 2.36E+00 1.21 9.50E-04 1.26 4.50E-03 1.26

1967 3.76E+04 1.34 1.75E+00 1.62 9.99E-04 1.26 4.18E-03 1.26

1968 4.59E+04 1.32 1.91E+00 1.58 6.21E-04 1.26 6.26E-03 1.26

1969 3.02E+04 1.26 2.92E+00 1.54 1.51E-01 1.26 1.58E-02 1.26

1970 3.16E+04 1.24 2.53E+00 1.33 2.83E-03 1.26 3.69E-03 1.26

1971 4.25E+04 1.24 1.74E+00 1.23 2.84E-03 1.26 1.19E-03 1.26

1972 5.26E+04 1.23 2.18E-01 1.53 1.96E-03 1.26 1.75E-03 1.26

1973 3.54E+04 1.23 1.36E-01 1.30 2.27E-03 1.26 1.15E-03 1.26

1974 5.79E+04 1.53 1.33E-01 1.16 7.24E-04 1.26 1.88E-03 1.26

1975 3.19E+04 1.40 8.62E-03 1.18 2.40E-04 1.26 1.03E-03 1.26

1976 1.67E+04 1.25 1.08E-02 1.17 1.58E-03 1.26 1.08E-03 1.26

1977 2.31E+04 1.25 4.40E-03 1.25 4.93E-04 1.26 3.45E-04 1.26

1978 2.13E+04 1.28 4.57E-03 1.18 6.91E-04 1.26 6.78E-04 1.26

1979 2.00E+04 1.25 6.31E-03 1.38 1.65E-04 1.26 5.39E-04 1.26

1980 1.79E+04 1.24 2.14E-03 1.64 3.81E-04 1.26 8.87E-04 1.26

1981 2.36E+04 1.25 3.71E-03 1.49 7.01E-04 1.26 1.37E-03 1.26

1982 2.71E+04 1.23 8.46E-03 1.51 5.11E-04 1.26 2.00E-03 1.26

1983 3.99E+04 1.24 6.38E-03 1.39 3.12E-04 1.26 9.80E-04 1.26

1984 5.02E+04 1.24 2.04E-02 1.29 1.65E-04 1.26 4.86E-04 1.26

1985 4.08E+04 1.25 4.90E-03 1.51 9.28E-05 1.26 5.59E-04 1.26

1986 2.58E+04 1.23 2.00E-03 1.37 2.16E-04 1.26 3.25E-04 1.26

1987 3.86E+04 1.32 1.63E-03 2.20 2.12E-04 1.26 1.85E-03 1.26

1988 2.40E+04 1.24 8.01E-05 2.45 1.03E-04 1.26 3.21E-04 1.26

1989 1.75E+04 1.19 2.52E-05 1.15 1.36E-04 1.26 1.10E-03 1.26

1990 1.63E+04 1.19 8.35E-04 1.99 2.24E-04 1.82 3.55E-04 1.82

1991 1.21E+04 1.41 6.55E-04 1.99 5.02E-04 1.82 1.84E-04 1.82

1992 1.08E+04 1.21 2.96E-06 1.99 5.93E-05 1.82 9.85E-05 1.82

1993 1.22E+04 1.97 3.44E-04 1.99 1.32E-04 1.82 1.20E-04 1.82

1994 1.05E+04 1.64 1.58E-06 1.99 1.46E-04 1.82 1.50E-04 1.82

1995 6.43E+03 1.49 3.16E-04 1.99 1.03E-04 1.82 9.87E-05 1.82

1996 3.44E+03 1.55 6.84E-04 1.99 4.74E-05 1.82 1.07E-04 1.82

1997 3.91E+03 1.50 1.65E-04 1.99 4.17E-06 1.82 7.00E-06 1.82

1998 4.34E+03 1.46 1.35E-06 1.99 1.23E-05 1.82 3.07E-06 1.82

1999 2.29E+02 2.41 1.35E-06 1.99 4.66E-05 1.82 3.21E-06 1.82

2000 1.97E+02 2.33 1.29E-03 1.99 4.57E-05 1.82 6.75E-06 1.82

2001 2.15E+02 1.98 1.55E-03 1.99 9.91E-06 1.82 8.62E-06 1.82

(a) Values are for plant perimeter.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 156 of 188

Table C-10 Central Shops Annual Intakes and geometric standard deviations, (Bq per 

year)

Tritium Iodine-131 Plutonium Uranium

Year 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD

1955 4.45E+04 1.21 1.49E+02 1.81 1.31E+00 1.82 1.51E-01 1.82

1956 9.48E+04 1.19 3.57E+03 1.67 5.06E-03 1.82 5.83E-02 1.82

1957 2.49E+05 1.21 7.28E+02 1.90 6.79E-03 1.82 9.76E-04 1.82

1958 3.27E+05 1.20 5.29E+01 2.31 3.32E-03 1.26 6.97E-04 1.26

1959 2.37E+05 1.17 4.06E+02 1.89 3.79E-03 1.26 3.09E-03 1.26

1960 1.97E+05 1.25 1.92E+01 2.29 1.13E-02 1.26 8.65E-03 1.26

1961 1.60E+05 1.24 2.08E+02 1.84 1.88E-03 1.26 4.42E-03 1.26

1962 1.73E+05 1.29 1.05E+01 3.09 1.57E-03 1.26 3.13E-03 1.26

1963 1.82E+05 1.35 6.03E+00 3.34 4.91E-04 1.26 1.11E-02 1.26

1964 2.42E+05 1.37 1.83E+00 1.99 7.86E-04 1.26 1.96E-02 1.26

1965 1.76E+05 1.34 2.47E+00 1.59 2.34E-03 1.26 1.82E-02 1.26

1966 1.93E+05 1.26 4.21E+00 1.43 1.43E-03 1.26 7.26E-03 1.26

1967 2.00E+05 1.25 3.49E+00 1.95 1.50E-03 1.26 6.75E-03 1.26

1968 1.83E+05 1.29 3.58E+00 1.87 9.44E-04 1.26 9.51E-03 1.26

1969 1.44E+05 1.23 5.35E+00 1.77 2.22E-01 1.26 2.44E-02 1.26

1970 1.32E+05 1.22 4.79E+00 1.68 4.22E-03 1.26 5.88E-03 1.26

1971 1.52E+05 1.24 3.10E+00 1.47 4.21E-03 1.26 1.84E-03 1.26

1972 1.65E+05 1.23 5.26E-01 2.19 2.90E-03 1.26 2.74E-03 1.26

1973 1.26E+05 1.24 2.39E-01 1.49 3.35E-03 1.26 1.81E-03 1.26

1974 1.54E+05 1.42 2.18E-01 1.20 1.09E-03 1.26 3.05E-03 1.26

1975 1.01E+05 1.31 1.40E-02 1.19 3.55E-04 1.26 1.66E-03 1.26

1976 6.63E+04 1.24 1.80E-02 1.24 2.34E-03 1.26 1.75E-03 1.26

1977 9.84E+04 1.27 8.32E-03 1.59 7.29E-04 1.26 5.48E-04 1.26

1978 8.56E+04 1.27 7.53E-03 1.23 1.02E-03 1.26 1.08E-03 1.26

1979 7.71E+04 1.29 1.07E-02 1.50 2.51E-04 1.26 8.60E-04 1.26

1980 7.49E+04 1.26 3.91E-03 1.85 5.68E-04 1.26 1.42E-03 1.26

1981 6.95E+04 1.25 7.63E-03 1.92 1.05E-03 1.26 2.15E-03 1.26

1982 8.84E+04 1.27 1.56E-02 1.75 8.02E-04 1.26 3.22E-03 1.26

1983 1.07E+05 1.26 1.17E-02 1.64 4.73E-04 1.26 1.58E-03 1.26

1984 1.53E+05 1.27 3.36E-02 1.33 2.57E-04 1.26 7.73E-04 1.26

1985 1.64E+05 1.27 8.24E-03 1.59 1.46E-04 1.26 8.91E-04 1.26

1986 6.73E+04 1.25 3.42E-03 1.49 3.23E-04 1.26 5.19E-04 1.26

1987 6.34E+04 1.32 2.97E-03 2.31 3.17E-04 1.26 2.99E-03 1.26

1988 4.28E+04 1.25 1.82E-04 2.67 1.62E-04 1.26 5.16E-04 1.26

1989 3.19E+04 1.18 4.08E-05 1.15 2.16E-04 1.26 1.78E-03 1.26

1990 2.98E+04 1.17 1.35E-03 1.99 3.45E-04 1.82 5.47E-04 1.82

1991 2.31E+04 1.33 1.06E-03 1.99 7.74E-04 1.82 2.84E-04 1.82

1992 1.84E+04 1.20 4.35E-06 1.99 9.10E-05 1.82 1.59E-04 1.82

1993 2.21E+04 2.34 5.31E-04 1.99 2.02E-04 1.82 1.93E-04 1.82

1994 1.76E+04 1.82 2.32E-06 1.99 2.20E-04 1.82 2.42E-04 1.82

1995 1.17E+04 1.51 4.94E-04 1.99 1.62E-04 1.82 1.60E-04 1.82

1996 7.00E+03 1.58 1.09E-03 1.99 7.29E-05 1.82 1.73E-04 1.82

1997 7.81E+03 1.50 2.43E-04 1.99 6.33E-06 1.82 1.05E-05 1.82

1998 8.46E+03 1.44 2.20E-06 1.99 1.86E-05 1.82 4.95E-06 1.82

1999 1.62E+03 1.73 2.20E-06 1.99 6.95E-05 1.82 5.14E-06 1.82

2000 1.81E+03 1.60 1.92E-03 1.99 6.86E-05 1.82 1.09E-05 1.82

2001 1.83E+03 1.46 2.32E-03 1.99 1.47E-05 1.82 1.40E-05 1.82
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 157 of 188

Table C-11 L Reactor Area Annual Intakes and geometric standard deviations, (Bq per 

year)

Tritium Iodine-131 Plutonium Uranium

Year 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD

1955 5.72E+03 2.71 9.81E+01 1.90 8.19E-01 1.82 9.94E-02 1.82

1956 4.07E+04 1.77 2.35E+03 1.68 3.15E-03 1.82 3.61E-02 1.82

1957 1.54E+05 1.64 5.25E+02 2.02 4.23E-03 1.82 6.08E-04 1.82

1958 2.11E+05 1.72 6.89E+01 2.94 2.06E-03 1.26 4.47E-04 1.26

1959 1.18E+05 2.09 2.84E+02 2.10 2.36E-03 1.26 1.91E-03 1.26

1960 1.30E+05 2.29 2.25E+01 2.93 6.95E-03 1.26 5.34E-03 1.26

1961 1.23E+05 2.44 1.51E+02 2.10 1.19E-03 1.26 2.76E-03 1.26

1962 1.66E+05 2.49 2.49E+01 3.35 1.01E-03 1.26 1.95E-03 1.26

1963 1.44E+05 2.80 1.54E+01 3.48 3.09E-04 1.26 6.99E-03 1.26

1964 3.14E+05 3.12 1.43E+01 3.07 5.15E-04 1.26 1.24E-02 1.26

1965 2.00E+05 2.98 7.72E+00 2.90 1.56E-03 1.26 1.13E-02 1.26

1966 1.63E+05 2.96 4.42E+00 2.20 9.50E-04 1.26 4.50E-03 1.26

1967 1.58E+05 3.03 1.28E+01 2.94 9.99E-04 1.26 4.18E-03 1.26

1968 1.33E+05 3.18 3.73E+00 2.52 6.21E-04 1.26 6.26E-03 1.26

1969 6.99E+04 1.29 4.96E+00 2.30 1.51E-01 1.26 1.58E-02 1.26

1970 8.06E+04 1.28 3.00E+00 1.71 2.83E-03 1.26 3.69E-03 1.26

1971 9.45E+04 1.27 1.81E+00 1.34 2.84E-03 1.26 1.19E-03 1.26

1972 9.96E+04 1.25 2.75E-01 1.96 1.96E-03 1.26 1.75E-03 1.26

1973 8.11E+04 1.27 1.64E-01 1.70 2.27E-03 1.26 1.15E-03 1.26

1974 9.95E+04 1.43 1.34E-01 1.18 7.24E-04 1.26 1.88E-03 1.26

1975 6.56E+04 1.35 8.96E-03 1.28 2.40E-04 1.26 1.03E-03 1.26

1976 4.58E+04 1.27 1.09E-02 1.19 1.58E-03 1.26 1.08E-03 1.26

1977 4.88E+04 1.28 4.64E-03 1.38 4.93E-04 1.26 3.45E-04 1.26

1978 4.50E+04 1.29 4.74E-03 1.29 6.91E-04 1.26 6.78E-04 1.26

1979 4.59E+04 1.28 8.06E-03 1.90 1.65E-04 1.26 5.39E-04 1.26

1980 4.30E+04 1.27 3.65E-03 2.35 3.81E-04 1.26 8.87E-04 1.26

1981 4.67E+04 1.27 5.02E-03 2.01 7.01E-04 1.26 1.37E-03 1.26

1982 5.25E+04 1.27 1.26E-02 2.14 5.11E-04 1.26 2.00E-03 1.26

1983 7.29E+04 1.31 8.15E-03 1.86 3.12E-04 1.26 9.80E-04 1.26

1984 8.36E+04 1.27 2.42E-02 1.68 1.65E-04 1.26 4.86E-04 1.26

1985 7.63E+04 1.30 7.79E-03 2.18 9.28E-05 1.26 5.59E-04 1.26

1986 5.07E+04 1.72 2.80E-03 1.95 2.16E-04 1.26 3.25E-04 1.26

1987 6.11E+04 1.64 6.65E-03 2.90 2.12E-04 1.26 1.85E-03 1.26

1988 4.57E+04 1.70 3.42E-03 3.16 1.03E-04 1.26 3.21E-04 1.26

1989 3.72E+04 1.98 2.52E-05 1.15 1.36E-04 1.26 1.10E-03 1.26

1990 3.20E+04 2.00 8.35E-04 1.99 2.24E-04 1.82 3.55E-04 1.82

1991 2.38E+04 2.20 6.55E-04 1.99 5.02E-04 1.82 1.84E-04 1.82

1992 1.90E+04 1.82 2.96E-06 1.99 5.93E-05 1.82 9.85E-05 1.82

1993 1.94E+04 3.54 3.44E-04 1.99 1.32E-04 1.82 1.20E-04 1.82

1994 1.45E+04 2.78 1.58E-06 1.99 1.46E-04 1.82 1.50E-04 1.82

1995 8.48E+03 2.33 3.16E-04 1.99 1.03E-04 1.82 9.87E-05 1.82

1996 5.43E+03 2.88 6.84E-04 1.99 4.74E-05 1.82 1.07E-04 1.82

1997 5.29E+03 2.41 1.65E-04 1.99 4.17E-06 1.82 7.00E-06 1.82

1998 5.48E+03 2.05 1.35E-06 1.99 1.23E-05 1.82 3.07E-06 1.82

1999 1.43E+03 3.27 1.35E-06 1.99 4.66E-05 1.82 3.21E-06 1.82

2000 1.54E+03 3.37 1.29E-03 1.99 4.57E-05 1.82 6.75E-06 1.82

2001 1.61E+03 2.49 1.55E-03 1.99 9.91E-06 1.82 8.62E-06 1.82
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 158 of 188

Table C-12 P Reactor Area Annual Intakes and geometric standard deviations, (Bq per 

year)

Tritium Iodine-131 Plutonium Uranium

Year 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD

1955 1.73E+04 1.43 8.85E+01 1.91 7.04E-01 1.82 9.26E-02 1.82

1956 1.01E+05 1.31 2.19E+03 1.69 2.67E-03 1.82 3.04E-02 1.82

1957 2.23E+05 1.27 5.25E+02 2.05 3.60E-03 1.82 5.18E-04 1.82

1958 2.74E+05 1.26 5.73E+01 2.82 1.75E-03 1.26 3.99E-04 1.26

1959 2.49E+05 1.31 2.47E+02 2.13 2.01E-03 1.26 1.60E-03 1.26

1960 4.08E+05 1.32 2.35E+01 3.06 5.79E-03 1.26 4.45E-03 1.26

1961 4.11E+05 1.32 1.16E+02 1.99 1.03E-03 1.26 2.35E-03 1.26

1962 6.30E+05 1.32 2.35E+02 3.43 9.18E-04 1.26 1.66E-03 1.26

1963 4.36E+05 1.37 1.44E+01 3.48 2.66E-04 1.26 6.03E-03 1.26

1964 6.41E+05 1.37 1.80E+00 2.59 4.79E-04 1.26 1.08E-02 1.26

1965 5.07E+05 1.34 1.69E+00 2.05 1.50E-03 1.26 9.39E-03 1.26

1966 3.70E+05 1.34 2.16E+00 1.44 9.05E-04 1.26 3.79E-03 1.26

1967 3.40E+05 1.35 2.50E+00 2.32 9.49E-04 1.26 3.51E-03 1.26

1968 2.30E+05 1.37 5.55E+00 2.88 5.81E-04 1.26 5.85E-03 1.26

1969 2.55E+05 1.29 9.40E+00 2.82 1.49E-01 1.26 1.44E-02 1.26

1970 3.24E+05 1.30 2.50E+00 1.71 2.74E-03 1.26 3.17E-03 1.26

1971 3.17E+05 1.28 1.80E+00 1.72 2.75E-03 1.26 1.08E-03 1.26

1972 3.39E+05 1.27 4.00E-01 2.56 1.92E-03 1.26 1.55E-03 1.26

1973 4.46E+05 1.26 2.63E-01 2.47 2.23E-03 1.26 1.01E-03 1.26

1974 4.48E+05 1.32 1.11E-01 1.17 6.89E-04 1.26 1.57E-03 1.26

1975 4.32E+05 1.29 7.32E-03 1.22 2.35E-04 1.26 8.54E-04 1.26

1976 3.25E+05 1.26 9.03E-03 1.18 1.54E-03 1.26 9.03E-04 1.26

1977 2.42E+05 1.29 3.79E-03 1.33 4.81E-04 1.26 2.99E-04 1.26

1978 2.20E+05 1.29 3.87E-03 1.23 6.76E-04 1.26 5.83E-04 1.26

1979 2.69E+05 1.29 3.33E-02 3.05 1.53E-04 1.26 4.63E-04 1.26

1980 2.08E+05 1.28 2.51E-02 3.08 3.68E-04 1.26 7.60E-04 1.26

1981 1.81E+05 1.27 2.05E-02 2.96 6.75E-04 1.26 1.20E-03 1.26

1982 1.75E+05 1.27 6.74E-02 3.01 4.51E-04 1.26 1.69E-03 1.26

1983 2.16E+05 1.29 3.18E-02 2.89 2.94E-04 1.26 8.24E-04 1.26

1984 3.02E+05 1.30 7.85E-02 2.90 1.48E-04 1.26 4.19E-04 1.26

1985 2.64E+05 1.29 4.94E-02 2.97 8.15E-05 1.26 4.80E-04 1.26

1986 2.65E+05 1.28 6.56E-03 2.69 2.09E-04 1.26 2.79E-04 1.26

1987 3.29E+05 1.29 4.99E-02 3.08 2.04E-04 1.26 1.54E-03 1.26

1988 3.07E+05 1.29 2.73E-04 3.01 9.02E-05 1.26 2.71E-04 1.26

1989 3.91E+05 1.06 2.09E-05 1.15 1.17E-04 1.26 9.20E-04 1.26

1990 2.27E+05 1.07 6.94E-04 1.99 2.04E-04 1.82 3.24E-04 1.82

1991 5.99E+04 1.15 5.45E-04 1.99 4.59E-04 1.82 1.68E-04 1.82

1992 1.41E+04 1.21 2.92E-06 1.99 5.46E-05 1.82 8.28E-05 1.82

1993 1.89E+04 2.31 3.15E-04 1.99 1.21E-04 1.82 1.01E-04 1.82

1994 4.66E+04 1.56 1.56E-06 1.99 1.39E-04 1.82 1.27E-04 1.82

1995 3.41E+04 1.46 2.82E-04 1.99 9.10E-05 1.82 8.23E-05 1.82

1996 2.67E+04 1.47 5.87E-04 1.99 4.35E-05 1.82 8.88E-05 1.82

1997 1.18E+04 1.49 1.63E-04 1.99 3.91E-06 1.82 6.63E-06 1.82

1998 1.45E+04 1.44 1.13E-06 1.99 1.16E-05 1.82 2.57E-06 1.82

1999 3.24E+02 3.18 1.13E-06 1.99 4.48E-05 1.82 2.73E-06 1.82

2000 2.90E+02 3.08 1.25E-03 1.99 4.37E-05 1.82 5.70E-06 1.82

2001 2.91E+02 2.57 1.48E-03 1.99 9.68E-06 1.82 7.18E-06 1.82
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 159 of 188

Table C-13 K Reactor Area Annual Intakes and geometric standard deviations, (Bq per 

year)

Tritium Iodine-131 Plutonium Uranium

Year 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD

1955 1.24E+04 1.46 1.09E+02 1.88 9.56E-01 1.82 1.06E-01 1.82

1956 6.46E+04 1.32 2.51E+03 1.68 3.73E-03 1.82 4.31E-02 1.82

1957 2.54E+05 1.27 5.26E+02 2.03 5.00E-03 1.82 7.18E-04 1.82

1958 2.67E+05 1.25 6.93E+01 2.95 2.45E-03 1.26 5.00E-04 1.26

1959 3.03E+05 1.31 3.26E+02 2.07 2.79E-03 1.26 2.30E-03 1.26

1960 4.56E+05 1.30 3.11E+01 3.01 8.37E-03 1.26 6.43E-03 1.26

1961 3.11E+05 1.31 1.73E+02 2.03 1.36E-03 1.26 3.25E-03 1.26

1962 5.37E+05 1.32 1.84E+01 3.29 1.12E-03 1.26 2.30E-03 1.26

1963 1.57E+05 3.30 7.31E+00 3.41 3.59E-04 1.26 8.13E-03 1.26

1964 7.72E+05 1.37 2.32E+00 2.51 5.51E-04 1.26 1.43E-02 1.26

1965 8.84E+05 1.32 3.41E+00 2.40 1.61E-03 1.26 1.36E-02 1.26

1966 6.70E+05 1.32 3.55E+00 1.68 9.83E-04 1.26 5.37E-03 1.26

1967 8.13E+05 1.34 1.43E+01 2.93 1.04E-03 1.26 5.00E-03 1.26

1968 7.13E+05 1.32 1.65E+01 3.10 6.58E-04 1.26 6.63E-03 1.26

1969 6.13E+05 1.30 2.67E+01 3.00 1.49E-01 1.26 1.73E-02 1.26

1970 7.64E+05 1.30 8.19E+00 2.61 2.87E-03 1.26 4.30E-03 1.26

1971 8.22E+05 1.29 2.21E+00 1.37 2.86E-03 1.26 1.30E-03 1.26

1972 7.37E+05 1.27 5.26E-01 2.48 1.96E-03 1.26 1.97E-03 1.26

1973 6.52E+05 1.28 4.09E-01 2.52 2.26E-03 1.26 1.31E-03 1.26

1974 5.39E+05 1.31 1.63E-01 1.21 7.51E-04 1.26 2.27E-03 1.26

1975 3.92E+05 1.31 1.51E-02 1.92 2.40E-04 1.26 1.24E-03 1.26

1976 4.12E+05 1.25 1.32E-02 1.21 1.59E-03 1.26 1.30E-03 1.26

1977 2.72E+05 1.28 5.98E-03 1.53 4.94E-04 1.26 4.00E-04 1.26

1978 2.84E+05 1.29 7.81E-03 1.89 6.89E-04 1.26 7.90E-04 1.26

1979 2.87E+05 1.29 8.71E-03 1.69 1.75E-04 1.26 6.30E-04 1.26

1980 3.37E+05 1.27 4.37E-03 2.35 3.88E-04 1.26 1.04E-03 1.26

1981 3.76E+05 1.27 5.41E-03 1.85 7.15E-04 1.26 1.55E-03 1.26

1982 3.72E+05 1.28 1.69E-02 2.25 5.79E-04 1.26 2.38E-03 1.26

1983 5.31E+05 1.31 8.61E-03 1.63 3.28E-04 1.26 1.17E-03 1.26

1984 3.82E+05 1.28 2.73E-02 1.54 1.84E-04 1.26 5.65E-04 1.26

1985 3.60E+05 1.28 9.56E-03 2.20 1.06E-04 1.26 6.53E-04 1.26

1986 2.65E+05 1.29 9.86E-03 2.70 2.20E-04 1.26 3.81E-04 1.26

1987 2.37E+05 1.29 5.24E-03 2.79 2.17E-04 1.26 2.22E-03 1.26

1988 2.45E+05 1.29 3.25E-04 2.97 1.18E-04 1.26 3.81E-04 1.26

1989 1.17E+05 1.09 3.04E-05 1.15 1.58E-04 1.26 1.32E-03 1.26

1990 1.63E+05 1.01 1.01E-03 1.99 2.44E-04 1.82 3.87E-04 1.82

1991 2.08E+05 0.98 7.91E-04 1.99 5.48E-04 1.82 2.01E-04 1.82

1992 1.72E+05 1.07 2.92E-06 1.99 6.41E-05 1.82 1.18E-04 1.82

1993 1.55E+05 5.03 3.76E-04 1.99 1.43E-04 1.82 1.43E-04 1.82

1994 5.83E+04 4.68 1.56E-06 1.99 1.52E-04 1.82 1.79E-04 1.82

1995 1.65E+04 3.88 3.54E-04 1.99 1.17E-04 1.82 1.19E-04 1.82

1996 1.31E+04 4.27 8.00E-04 1.99 5.14E-05 1.82 1.29E-04 1.82

1997 1.17E+04 3.93 1.63E-04 1.99 4.40E-06 1.82 7.29E-06 1.82

1998 1.00E+04 3.58 1.64E-06 1.99 1.29E-05 1.82 3.67E-06 1.82

1999 7.01E+03 4.32 1.64E-06 1.99 4.75E-05 1.82 3.78E-06 1.82

2000 5.02E+03 4.40 1.31E-03 1.99 4.72E-05 1.82 8.01E-06 1.82

2001 3.32E+03 4.19 1.59E-03 1.99 9.93E-06 1.82 1.04E-05 1.82
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 160 of 188

Table C-14 R Reactor Area Annual Intakes and geometric standard deviations, (Bq per 

year)

Tritium Iodine-131 Plutonium Uranium

Year 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD

1955 4.12E+03 2.09 1.31E+02 1.86 9.76E-01 1.82 1.46E-01 1.82

1956 4.14E+04 1.75 3.44E+03 1.68 3.61E-03 1.82 4.05E-02 1.82

1957 2.05E+05 1.44 8.55E+02 1.98 4.89E-03 1.82 7.03E-04 1.82

1958 2.89E+05 1.57 7.54E+01 2.58 2.36E-03 1.26 5.92E-04 1.26

1959 1.49E+05 1.94 3.26E+02 2.05 2.75E-03 1.26 2.09E-03 1.26

1960 1.47E+05 2.27 3.55E+01 3.13 7.55E-03 1.26 5.82E-03 1.26

1961 1.20E+05 1.84 1.65E+02 2.12 1.47E-03 1.26 3.20E-03 1.26

1962 1.73E+05 2.40 2.58E+01 3.34 1.39E-03 1.26 2.26E-03 1.26

1963 1.63E+05 2.37 1.01E+02 3.51 3.71E-04 1.26 8.41E-03 1.26

1964 2.16E+05 2.46 1.72E+00 2.37 7.54E-04 1.26 1.54E-02 1.26

1965 7.96E+04 1.40 1.66E+00 1.60 2.47E-03 1.26 1.23E-02 1.26

1966 6.91E+04 1.39 2.63E+00 1.27 1.48E-03 1.26 5.03E-03 1.26

1967 7.09E+04 1.38 2.24E+00 1.90 1.54E-03 1.26 4.65E-03 1.26

1968 8.24E+04 1.35 2.39E+00 1.87 9.25E-04 1.26 9.31E-03 1.26

1969 5.65E+04 1.26 3.69E+00 1.83 2.55E-01 1.26 2.19E-02 1.26

1970 6.08E+04 1.25 2.84E+00 1.42 4.58E-03 1.26 4.41E-03 1.26

1971 7.88E+04 1.24 1.92E+00 1.27 4.61E-03 1.26 1.65E-03 1.26

1972 9.58E+04 1.23 2.57E-01 1.71 3.25E-03 1.26 2.26E-03 1.26

1973 6.93E+04 1.24 1.57E-01 1.45 3.81E-03 1.26 1.47E-03 1.26

1974 1.07E+05 1.51 1.45E-01 1.17 1.12E-03 1.26 2.04E-03 1.26

1975 6.30E+04 1.38 9.52E-03 1.19 4.00E-04 1.26 1.11E-03 1.26

1976 3.50E+04 1.26 1.17E-02 1.18 2.59E-03 1.26 1.18E-03 1.26

1977 4.39E+04 1.26 4.88E-03 1.29 8.12E-04 1.26 4.21E-04 1.26

1978 4.04E+04 1.28 4.99E-03 1.20 1.15E-03 1.26 8.10E-04 1.26

1979 3.95E+04 1.25 8.11E-03 1.75 2.43E-04 1.26 6.40E-04 1.26

1980 3.45E+04 1.25 3.36E-03 2.18 6.12E-04 1.26 1.05E-03 1.26

1981 4.37E+04 1.25 4.86E-03 1.85 1.12E-03 1.26 1.74E-03 1.26

1982 4.92E+04 1.24 1.19E-02 1.96 6.56E-04 1.26 2.26E-03 1.26

1983 7.19E+04 1.25 8.12E-03 1.72 4.72E-04 1.26 1.09E-03 1.26

1984 9.04E+04 1.24 2.48E-02 1.57 2.21E-04 1.26 5.83E-04 1.26

1985 7.36E+04 1.25 7.28E-03 2.01 1.17E-04 1.26 6.63E-04 1.26

1986 4.97E+04 1.24 2.45E-03 1.62 3.47E-04 1.26 3.85E-04 1.26

1987 7.25E+04 1.32 3.90E-03 2.72 3.38E-04 1.26 2.01E-03 1.26

1988 4.76E+04 1.25 1.54E-04 2.78 1.29E-04 1.26 3.63E-04 1.26

1989 3.82E+04 1.17 2.72E-05 1.15 1.64E-04 1.26 1.21E-03 1.26

1990 3.27E+04 1.18 9.02E-04 1.99 3.14E-04 1.82 4.98E-04 1.82

1991 2.19E+04 1.40 7.08E-04 1.99 7.05E-04 1.82 2.58E-04 1.82

1992 1.85E+04 1.21 5.01E-06 1.99 8.49E-05 1.82 1.10E-04 1.82

1993 2.10E+04 2.03 4.83E-04 1.99 1.87E-04 1.82 1.34E-04 1.82

1994 1.86E+04 1.66 2.67E-06 1.99 2.25E-04 1.82 1.70E-04 1.82

1995 1.15E+04 1.50 4.17E-04 1.99 1.32E-04 1.82 1.07E-04 1.82

1996 6.34E+03 1.57 8.10E-04 1.99 6.72E-05 1.82 1.15E-04 1.82

1997 6.79E+03 1.51 2.79E-04 1.99 6.25E-06 1.82 1.08E-05 1.82

1998 7.56E+03 1.47 1.46E-06 1.99 1.86E-05 1.82 3.38E-06 1.82

1999 3.99E+02 2.51 1.46E-06 1.99 7.44E-05 1.82 3.72E-06 1.82

2000 3.39E+02 2.46 2.08E-03 1.99 7.16E-05 1.82 7.66E-06 1.82

2001 3.67E+02 2.07 2.45E-03 1.99 1.63E-05 1.82 9.37E-06 1.82
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 161 of 188

Table C-15 Rail Yard Annual Intakes and geometric standard deviations, (Bq per year)

Tritium Iodine-131 Plutonium Uranium

Year 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD

1955 2.80E+03 1.53 6.67E+01 1.82 5.47E-01 1.82 7.13E-02 1.82

1956 2.55E+04 1.32 1.68E+03 1.67 2.08E-03 1.82 2.37E-02 1.82

1957 1.01E+05 1.27 3.74E+02 1.92 2.80E-03 1.82 4.02E-04 1.82

1958 1.35E+05 1.26 2.84E+01 2.37 1.36E-03 1.26 3.09E-04 1.26

1959 6.79E+04 1.34 1.72E+02 1.92 1.56E-03 1.26 1.24E-03 1.26

1960 7.16E+04 1.39 9.78E+00 2.60 4.51E-03 1.26 3.47E-03 1.26

1961 6.79E+04 1.36 8.49E+01 1.88 8.02E-04 1.26 1.83E-03 1.26

1962 9.11E+04 1.42 2.74E+01 3.38 7.09E-04 1.26 1.29E-03 1.26

1963 7.50E+04 1.51 6.58E+00 3.45 2.07E-04 1.26 4.68E-03 1.26

1964 1.09E+05 1.55 1.05E+00 2.38 3.69E-04 1.26 8.39E-03 1.26

1965 6.82E+04 1.46 1.12E+00 1.83 1.15E-03 1.26 7.31E-03 1.26

1966 5.45E+04 1.46 1.62E+00 1.35 6.95E-04 1.26 2.94E-03 1.26

1967 5.50E+04 1.45 1.64E+00 2.15 7.29E-04 1.26 2.73E-03 1.26

1968 5.36E+04 1.43 1.83E+00 2.23 4.47E-04 1.26 4.50E-03 1.26

1969 4.23E+04 1.28 2.89E+00 2.19 1.14E-01 1.26 1.11E-02 1.26

1970 4.90E+04 1.28 1.77E+00 1.51 2.10E-03 1.26 2.46E-03 1.26

1971 5.70E+04 1.26 1.18E+00 1.35 2.11E-03 1.26 8.34E-04 1.26

1972 6.41E+04 1.25 1.74E-01 1.93 1.47E-03 1.26 1.20E-03 1.26

1973 5.74E+04 1.26 1.07E-01 1.72 1.71E-03 1.26 7.85E-04 1.26

1974 7.39E+04 1.44 8.65E-02 1.17 5.29E-04 1.26 1.22E-03 1.26

1975 5.33E+04 1.34 5.68E-03 1.21 1.80E-04 1.26 6.65E-04 1.26

1976 3.40E+04 1.27 7.02E-03 1.18 1.18E-03 1.26 7.03E-04 1.26

1977 3.49E+04 1.28 2.93E-03 1.32 3.69E-04 1.26 2.32E-04 1.26

1978 3.17E+04 1.29 3.00E-03 1.22 5.18E-04 1.26 4.53E-04 1.26

1979 3.45E+04 1.28 6.99E-03 2.30 1.18E-04 1.26 3.59E-04 1.26

1980 2.86E+04 1.27 3.81E-03 2.68 2.82E-04 1.26 5.90E-04 1.26

1981 3.03E+04 1.27 4.27E-03 2.32 5.18E-04 1.26 9.31E-04 1.26

1982 3.33E+04 1.26 1.18E-02 2.48 3.49E-04 1.26 1.31E-03 1.26

1983 4.58E+04 1.27 6.91E-03 2.20 2.26E-04 1.26 6.41E-04 1.26

1984 5.88E+04 1.27 1.96E-02 2.05 1.15E-04 1.26 3.25E-04 1.26

1985 5.04E+04 1.27 7.85E-03 2.52 6.31E-05 1.26 3.72E-04 1.26

1986 3.76E+04 1.27 1.89E-03 1.99 1.60E-04 1.26 2.17E-04 1.26

1987 4.99E+04 1.32 6.10E-03 2.96 1.57E-04 1.26 1.20E-03 1.26

1988 3.94E+04 1.28 1.53E-04 2.94 6.99E-05 1.26 2.11E-04 1.26

1989 3.99E+04 1.12 1.63E-05 1.15 9.10E-05 1.26 7.16E-04 1.26

1990 2.80E+04 1.13 5.41E-04 1.99 1.58E-04 1.82 2.50E-04 1.82

1991 1.39E+04 1.32 4.24E-04 1.99 3.54E-04 1.82 1.30E-04 1.82

1992 9.53E+03 1.21 2.24E-06 1.99 4.21E-05 1.82 6.44E-05 1.82

1993 1.07E+04 2.44 2.43E-04 1.99 9.32E-05 1.82 7.83E-05 1.82

1994 1.08E+04 1.80 1.19E-06 1.99 1.07E-04 1.82 9.85E-05 1.82

1995 6.81E+03 1.55 2.18E-04 1.99 7.04E-05 1.82 6.40E-05 1.82

1996 4.23E+03 1.63 4.55E-04 1.99 3.35E-05 1.82 6.91E-05 1.82

1997 3.53E+03 1.60 1.25E-04 1.99 3.01E-06 1.82 5.10E-06 1.82

1998 3.98E+03 1.53 8.77E-07 1.99 8.92E-06 1.82 2.00E-06 1.82

1999 2.27E+02 3.06 8.77E-07 1.99 3.44E-05 1.82 2.12E-06 1.82

2000 2.01E+02 2.96 9.56E-04 1.99 3.35E-05 1.82 4.43E-06 1.82

2001 2.04E+02 2.47 1.14E-03 1.99 7.42E-06 1.82 5.59E-06 1.82
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 162 of 188

Table C-16 Average Site-Wide (a) Annual Intakes and Geometric Standard Deviations, 

(Bq per year)

Tritium Iodine-131 Plutonium Uranium

Year 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD

1955 2.86E+04 1.33E+00 4.31E+02 1.80E+00 3.64E+00 1.82E+00 4.53E-01 1.82E+00

1956 1.61E+05 1.27E+00 1.06E+04 1.66E+00 1.40E-02 1.82E+00 1.60E-01 1.82E+00

1957 6.29E+05 1.24E+00 2.26E+03 1.89E+00 1.88E-02 1.82E+00 2.70E-03 1.82E+00

1958 8.42E+05 1.23E+00 1.51E+02 2.16E+00 9.15E-03 1.26E+00 2.01E-03 1.26E+00

1959 4.15E+05 1.26E+00 1.12E+03 1.87E+00 1.05E-02 1.26E+00 8.43E-03 1.26E+00

1960 3.73E+05 1.31E+00 5.04E+01 2.24E+00 3.06E-02 1.26E+00 2.35E-02 1.26E+00

1961 3.40E+05 1.28E+00 5.61E+02 1.83E+00 5.30E-03 1.26E+00 1.22E-02 1.26E+00

1962 4.18E+05 1.35E+00 3.20E+01 3.13E+00 4.58E-03 1.26E+00 8.65E-03 1.26E+00

1963 3.82E+05 1.43E+00 1.48E+01 3.32E+00 1.37E-03 1.26E+00 3.11E-02 1.26E+00

1964 5.54E+05 1.44E+00 4.76E+00 1.93E+00 2.35E-03 1.26E+00 5.55E-02 1.26E+00

1965 3.27E+05 1.38E+00 6.48E+00 1.52E+00 7.19E-03 1.26E+00 4.96E-02 1.26E+00

1966 3.01E+05 1.35E+00 1.06E+01 1.25E+00 4.36E-03 1.26E+00 1.99E-02 1.26E+00

1967 3.04E+05 1.35E+00 8.67E+00 1.82E+00 4.58E-03 1.26E+00 1.85E-02 1.26E+00

1968 3.26E+05 1.34E+00 9.19E+00 1.76E+00 2.83E-03 1.26E+00 2.85E-02 1.26E+00

1969 2.34E+05 1.25E+00 1.41E+01 1.72E+00 6.99E-01 1.26E+00 7.14E-02 1.26E+00

1970 2.48E+05 1.24E+00 1.15E+01 1.41E+00 1.31E-02 1.26E+00 1.64E-02 1.26E+00

1971 2.95E+05 1.24E+00 7.75E+00 1.26E+00 1.31E-02 1.26E+00 5.38E-03 1.26E+00

1972 3.40E+05 1.23E+00 1.02E+00 1.65E+00 9.08E-03 1.26E+00 7.84E-03 1.26E+00

1973 2.64E+05 1.23E+00 6.19E-01 1.38E+00 1.05E-02 1.26E+00 5.16E-03 1.26E+00

1974 3.52E+05 1.47E+00 5.87E-01 1.17E+00 3.32E-03 1.26E+00 8.30E-03 1.26E+00

1975 2.15E+05 1.35E+00 3.82E-02 1.19E+00 1.11E-03 1.26E+00 4.52E-03 1.26E+00

1976 1.44E+05 1.24E+00 4.77E-02 1.18E+00 7.29E-03 1.26E+00 4.77E-03 1.26E+00

1977 1.65E+05 1.25E+00 1.97E-02 1.29E+00 2.28E-03 1.26E+00 1.54E-03 1.26E+00

1978 1.53E+05 1.27E+00 2.03E-02 1.20E+00 3.20E-03 1.26E+00 3.02E-03 1.26E+00

1979 1.45E+05 1.26E+00 2.90E-02 1.48E+00 7.50E-04 1.26E+00 2.40E-03 1.26E+00

1980 1.38E+05 1.24E+00 1.05E-02 1.82E+00 1.76E-03 1.26E+00 3.94E-03 1.26E+00

1981 1.61E+05 1.24E+00 1.74E-02 1.62E+00 3.23E-03 1.26E+00 6.12E-03 1.26E+00

1982 1.78E+05 1.24E+00 4.02E-02 1.66E+00 2.29E-03 1.26E+00 8.86E-03 1.26E+00

1983 2.49E+05 1.25E+00 2.94E-02 1.49E+00 1.43E-03 1.26E+00 4.33E-03 1.26E+00

1984 3.07E+05 1.25E+00 9.25E-02 1.36E+00 7.44E-04 1.26E+00 2.16E-03 1.26E+00

1985 2.74E+05 1.25E+00 2.35E-02 1.67E+00 4.15E-04 1.26E+00 2.48E-03 1.26E+00

1986 1.74E+05 1.24E+00 9.31E-03 1.49E+00 9.97E-04 1.26E+00 1.45E-03 1.26E+00

1987 2.19E+05 1.31E+00 9.22E-03 2.42E+00 9.77E-04 1.26E+00 8.14E-03 1.26E+00

1988 1.44E+05 1.25E+00 4.91E-04 2.67E+00 4.61E-04 1.26E+00 1.42E-03 1.26E+00

1989 1.06E+05 1.18E+00 1.11E-04 1.15E+00 6.05E-04 1.26E+00 4.85E-03 1.26E+00

1990 9.48E+04 1.17E+00 3.68E-03 1.99E+00 1.01E-03 1.82E+00 1.61E-03 1.82E+00

1991 6.96E+04 1.35E+00 2.89E-03 1.99E+00 2.27E-03 1.82E+00 8.34E-04 1.82E+00

1992 5.85E+04 1.20E+00 1.37E-05 1.99E+00 2.69E-04 1.82E+00 4.35E-04 1.82E+00

1993 6.50E+04 2.46E+00 1.56E-03 1.99E+00 5.97E-04 1.82E+00 5.30E-04 1.82E+00

1994 5.36E+04 1.87E+00 7.34E-06 1.99E+00 6.70E-04 1.82E+00 6.64E-04 1.82E+00

1995 3.27E+04 1.57E+00 1.42E-03 1.99E+00 4.62E-04 1.82E+00 4.35E-04 1.82E+00

1996 1.84E+04 1.69E+00 3.04E-03 1.99E+00 2.15E-04 1.82E+00 4.70E-04 1.82E+00

1997 1.98E+04 1.60E+00 7.67E-04 1.99E+00 1.90E-05 1.82E+00 3.21E-05 1.82E+00

1998 2.18E+04 1.52E+00 5.97E-06 1.99E+00 5.63E-05 1.82E+00 1.35E-05 1.82E+00

1999 2.01E+03 2.54E+00 5.97E-06 1.99E+00 2.14E-04 1.82E+00 1.42E-05 1.82E+00

2000 1.91E+03 2.36E+00 5.94E-03 1.99E+00 2.10E-04 1.82E+00 2.99E-05 1.82E+00

2001 1.92E+03 1.98E+00 7.12E-03 1.99E+00 4.58E-05 1.82E+00 3.80E-05 1.82E+00

(a) Values are for plant perimeter.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 163 of 188

Table C-17 Maximum Site-Wide Annual Intakes and geometric standard deviations, (Bq 

per year)

Tritium Iodine-131 Plutonium Uranium

Year 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD 50th

Percentile

GSD

1955 2.12E+05 1.18 2.06E+03 1.79 2.17E+00 1.82 2.46E+01 1.82

1956 1.02E+06 1.26 5.03E+04 1.66 1.22E+00 1.82 1.02E-01 1.82

1957 2.88E+06 1.24 1.50E+04 1.87 1.94E-02 1.82 1.35E-01 1.82

1958 3.93E+06 1.20 9.41E+02 1.98 1.01E-02 1.26 6.74E-02 1.26

1959 2.12E+06 1.22 7.58E+03 1.84 6.73E-02 1.26 7.48E-02 1.26

1960 1.68E+06 1.22 3.57E+02 2.00 1.89E-01 1.26 2.47E-01 1.26

1961 1.72E+06 1.20 4.50E+03 1.80 8.75E-02 1.26 3.26E-02 1.26

1962 2.01E+06 1.24 2.35E+02 3.43 6.20E-02 1.26 2.06E-02 1.26

1963 1.97E+06 1.22 1.01E+02 3.51 2.05E-01 1.26 9.06E-03 1.26

1964 2.54E+06 1.28 2.69E+01 2.09 3.39E-01 1.26 1.10E-02 1.26

1965 1.00E+06 1.23 4.62E+01 1.18 3.99E-01 1.26 4.19E-02 1.26

1966 1.01E+06 1.23 8.21E+01 1.12 1.53E-01 1.26 2.45E-02 1.26

1967 9.22E+05 1.24 5.16E+01 1.34 1.43E-01 1.26 2.52E-02 1.26

1968 1.27E+06 1.22 5.91E+01 1.43 1.41E-01 1.26 1.40E-02 1.26

1969 8.03E+05 1.21 9.07E+01 1.42 3.07E-01 1.26 4.82E+00 1.26

1970 8.58E+05 1.21 8.60E+01 1.17 1.10E-01 1.26 8.10E-02 1.26

1971 1.10E+06 1.21 6.09E+01 1.16 2.31E-02 1.26 8.23E-02 1.26

1972 1.47E+06 1.20 6.88E+00 1.21 4.24E-02 1.26 5.97E-02 1.26

1973 1.00E+06 1.19 4.63E+00 1.17 2.91E-02 1.26 7.10E-02 1.26

1974 1.68E+06 1.53 4.72E+00 1.15 6.73E-02 1.26 1.84E-02 1.26

1975 8.69E+05 1.39 2.99E-01 1.16 3.67E-02 1.26 7.37E-03 1.26

1976 4.69E+05 1.20 3.85E-01 1.16 3.81E-02 1.26 4.66E-02 1.26

1977 5.77E+05 1.21 1.50E-01 1.17 9.85E-03 1.26 1.48E-02 1.26

1978 5.58E+05 1.24 1.62E-01 1.16 2.02E-02 1.26 2.11E-02 1.26

1979 5.06E+05 1.20 2.08E-01 1.17 1.63E-02 1.26 3.61E-03 1.26

1980 4.50E+05 1.19 6.24E-02 1.77 2.71E-02 1.26 1.07E-02 1.26

1981 7.13E+05 1.21 1.18E-01 1.19 3.49E-02 1.26 1.94E-02 1.26

1982 7.41E+05 1.20 2.65E-01 1.31 6.68E-02 1.26 1.27E-02 1.26

1983 1.16E+06 1.22 2.11E-01 1.17 3.35E-02 1.26 7.37E-03 1.26

1984 1.41E+06 1.21 7.01E-01 1.16 1.43E-02 1.26 3.63E-03 1.26

1985 1.05E+06 1.21 1.52E-01 1.50 1.69E-02 1.26 2.43E-03 1.26

1986 8.77E+05 1.19 6.60E-02 1.17 9.88E-03 1.26 6.06E-03 1.26

1987 1.28E+06 1.30 4.99E-02 3.08 6.55E-02 1.26 5.80E-03 1.26

1988 7.17E+05 1.21 3.42E-03 3.16 1.07E-02 1.26 2.72E-03 1.26

1989 4.38E+05 1.21 9.05E-04 1.15 3.84E-02 1.26 3.96E-03 1.26

1990 4.33E+05 1.20 3.00E-02 1.99 6.58E-03 1.82 4.15E-03 1.82

1991 3.28E+05 1.46 2.36E-02 1.99 3.42E-03 1.82 9.31E-03 1.82

1992 3.06E+05 1.21 9.47E-05 1.99 3.37E-03 1.82 1.18E-03 1.82

1993 3.54E+05 3.05 6.40E-03 1.99 4.09E-03 1.82 2.56E-03 1.82

1994 3.16E+05 1.68 5.06E-05 1.99 5.00E-03 1.82 3.66E-03 1.82

1995 1.95E+05 1.41 7.31E-03 1.99 3.52E-03 1.82 2.58E-03 1.82

1996 1.01E+05 1.42 2.08E-02 1.99 3.84E-03 1.82 9.11E-04 1.82

1997 1.21E+05 1.41 5.27E-03 1.99 1.74E-04 1.82 9.60E-05 1.82

1998 1.35E+05 1.41 4.88E-05 1.99 1.06E-04 1.82 2.92E-04 1.82

1999 7.01E+03 4.32 4.88E-05 1.99 1.02E-04 1.82 1.29E-03 1.82

2000 5.80E+03 4.03 3.66E-02 1.99 2.23E-04 1.82 1.20E-03 1.82

2001 5.78E+03 2.99 4.14E-02 1.99 3.08E-04 1.82 2.98E-04 1.82
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 164 of 188

C.1.2 Resuspension of Radioactive Materials from Soil

Soil analyses indicated that elevated levels of certain radionuclides. including plutonium, 

were detected at several locations near the release points in both F and H Areas. It was 

concluded that the measured soil concentrations in F and H Areas were due to 

deposition during the years of the greatest releases. Annual intakes based on 

ventilation rate of 2,400 cubic meters per year resulting from resuspension of 

radionuclides in soil are presented in Table C-18.

Table C-18 Annual Intakes for F and H Areas (Bq per year)

90Sr 137Cs 238Pu 239Pu

Location 50th GSD 50th GSD 50th GSD 50th GSD

F Area 2.46E-04 1.81 1.19E-02 1.06 4.59E-03 1.04 6.04E-03 1.03

H Area 3.29E-04 1.63 1.50E-02 1.04 2.42E-04 1.09 6.30E-04 1.05

For purposes of reconstructing potential missed or unmonitored dose, the annual intakes 

provided in Table C-18 should be used only for individuals whose employment records 

indicate that they spent considerable time in either F or H Areas. The 239Pu annual 

intakes should be assumed to have persisted since 1955 and the 238Pu annual intakes 

since 1969. Annual intakes of 90Sr and 137Cs should be conservatively assumed to have 

persisted essentially unchanged since 1955.

C.2 External Dose

C.2.1 Ambient Radiation

The ambient radiation levels in Table C-19 are presented in mrem/year for various 

locations at the SRP. These levels are for a full work year of 2,000 hours which 

represent a fraction (i.e., 0.228) of the full annual doses ( i.e., 8766 hours) that were 

derived from the SRS Annual Environmental Reports. The values in Table C-19 include 

the contributions from natural background radiation and nuclear weapons fallout. 

Therefore, use of these data to account for potential missed or unmonitored dose should 

in all cases represent an overestimation of the worker’s actual dose.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 165 of 188

Table C-19 Ambient radiation levels for selected SRS locations (mrem/year)

A Area C Area D-Area E Area F Area H Area S Area Z Area

Year 50th GSD 50th GSD 50th GSD 50th GSD 50th GSD 50th GSD 50th GSD 50th GSD

1954 73 62 58 62

1955 78 69 75 75 74

1956 75 74 69 77 90

1957 96 86 80 90 107

1958 102 103 98 98 120

1959 59 56 54 88 77

1960 61 41 40 182 81

1961 44 41 38 49 57

1962 64 47 50 49 61

1963 68 50 57 60 81

1964 35 48 30 58 75

1965 33 48 28 48 63

1966 24 22 49 105

1967 23 20 38 89

1968 18 18 40 48

1969 20 23 51 67

1970 19 13 56 81

1971 19 14 68 47

1972 22 17 12 76 47

1973 26 2.51 19 2.77 8 4.34 79 4.34 55 4.58

1974 22 1.19 18 1.23 23 1.77 168 1.10 128 1.10

1975 58 1.22 21 1.29 18 1.23 168 1.13 133 1.16

1976 128 1.17 23 1.26 18 1.24 140 1.17 142 1.12

1977 192 1.15 21 1.20 20 1.21 175 1.16 183 1.16

1978 43 1.19 21 1.20 21 1.20 100 1.17 83 1.18

1979 53 1.16 31 1.20 15 1.27 129 1.13 93 1.18

1980 34 1.18 18 1.23 17 1.25 80 1.08 56 1.19

1981 55 1.19 23 1.26 22 1.19 71 1.15 180 1.07

1982 42 1.10 28 1.22 20 1.21 27 1.23 65 1.19 142 1.16

1983 39 1.21 20 1.21 18 1.23 31 1.20 63 1.20 133 1.16

1984 27 1.23 23 1.19 21 1.20 39 1.21 70 1.18 125 1.17 22 1.19

1985 33 1.19 18 1.24 20 1.21 41 1.20 48 1.18 467 1.15 48 1.18 18 1.23

1986 37 1.23 21 1.20 22 1.19 29 1.21 60 1.18 117 1.16 46 1.18 18 1.23

1987 26 1.24 18 1.24 21 1.20 29 1.21 47 1.18 92 1.16 30 1.21 18 1.24

1988 27 1.44 20 1.40 18 1.43 27 1.44 43 1.41 167 1.41 27 1.44 19 1.41

1989 33 1.41 18 1.45 18 1.43 27 1.44 37 1.43 79 1.42 27 1.44 18 1.45

1990 21 1.42 12 1.41 16 1.42 182 1.43 34 1.43 55 1.42 23 1.43 13 1.42

1991 20 1.42 11 1.42 16 1.42 110 1.42 30 1.43 35 1.43 23 1.42 12 1.42

1992 18 1.48 10 1.11 13 1.11 76 1.11 24 1.11 29 1.11 20 1.11 13 1.39

1993 24 1.11 16 1.11 18 1.11 79 1.11 29 1.11 33 1.11 27 1.11 18 1.11

1994 24 1.11 17 1.11 18 1.11 78 1.11 31 1.11 36 1.11 28 1.11 19 1.11

1995 25 1.11 18 1.11 19 1.11 56 1.11 32 1.11 34 1.11 29 1.11 20 1.11

1996 26 1.11 19 1.11 19 1.11 55 1.11 31 1.11 34 1.11 28 1.11 19 1.11

1997 26 1.11 42 1.11 31 1.11 34 1.11 27 1.11 20 1.11

1998 27 1.11 35 1.11 33 1.11 33 1.11 29 1.11 22 1.11

1999 18 1.11

2000 19 1.11

2001 17 1.11
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 166 of 188

Table C-19 Continued

Forest

Service(a)

Central 

Shops

L-Area P Area K Area R Area Rail 

Yard

Site Perimeter

Year 50th GSD 50th GSD 50th GSD 50th GSD 50th GSD 50th GSD 50th GSD 50th GSD

1954 67 77

1955 84 83 80 75

1956 69 78 70 75 73 69

1957 83 95 92 91 93 83

1958 90 110 104 106 103 90

1959 41 61 56 55 64 41

1960 30 48 42 47 47 30

1961 26 38 36 43 33 26

1962 37 50 49 49 49 37

1963 36 59 58 57 58 36

1964 32 52 43 45 44 32

1965 28 28

1966 21 21

1967 19 19

1968 22 22

1969 22 22

1970 15 15

1971 14 14

1972 12 18 32 12

1973 17 1.25 22 2.73 35 ` 17 1.25

1974 16 1.26 23 1.35 55 1.22 16 1.26

1975 16 1.26 26 1.24 28 1.22 63 1.20 16 1.26

1976 18 1.23 23 1.26 22 1.36 54 1.19 18 1.23

1977 21 1.20 68 1.06 23 1.19 58 1.18 21 1.20

1978 23 1.19 133 1.16 22 1.19 52 1.20 23 1.19

1979 23 1.19 38 1.17 36 1.18 60 1.21 23 1.19

1980 22 1.19 20 1.30 18 1.23 38 1.17 22 1.19

1981 23 1.49 27 1.23 21 1.20 22 1.19 39 1.16 23 1.49

1982 18 1.23 23 1.19 20 1.21 29 1.21 45 1.19 18 1.23

1983 21 1.20 22 1.19 18 1.23 20 1.30 39 1.21 21 1.20

1984 20 1.21 65 1.19 19 1.41 22 1.19 43 1.20 19 1.22 23 1.19 20 1.21

1985 19 1.22 58 1.18 19 1.22 21 1.20 36 1.23 21 1.20 23 1.26 19 1.22

1986 19 1.22 29 1.21 24 1.26 21 1.20 40 1.21 22 1.19 23 1.19 19 1.22

1987 31 1.14 45 1.19 22 1.19 22 1.19 36 1.18 18 1.23 21 1.20 31 1.14

1988 21 1.47 100 1.41 22 1.45 19 1.41 36 1.44 20 1.40 23 1.44 21 1.47

1989 21 1.47 125 1.42 22 1.53 18 1.43 48 1.41 21 1.38 23 1.44 21 1.47

1990 24 1.43 116 1.42 15 1.43 14 1.41 28 1.42 15 1.43 16 1.43 24 1.43

1991 33 1.42 97 1.43 14 1.42 14 1.42 18 1.41 15 1.43 15 1.42 33 1.42

1992 14 1.11 112 1.11 13 1.11 13 1.11 17 1.11 13 1.11 13 1.11 14 1.11

1993 23 1.11 105 1.11 18 1.11 18 1.11 21 1.11 17 1.11 20 1.11 23 1.11

1994 21 1.11 62 1.11 19 1.11 18 1.11 21 1.11 18 1.11 20 1.11 21 1.11

1995 24 1.11 63 1.11 20 1.11 20 1.11 23 1.11 19 1.11 22 1.11 24 1.11

1996 23 1.11 57 1.11 19 1.11 19 1.11 20 1.11 19 1.11 22 1.11 23 1.11

1997 18 1.11 45 1.11 18 1.11 19 1.11 21 1.11 18 1.11 22 1.11 18 1.11

1998 17 1.11 45 1.11 22 1.11 22 1.11 23 1.11 21 1.11 23 1.11 17 1.11

1999 20 1.11 20 1.11

2000 21 1.11 21 1.11

2001 19 1.11 19 1.11

(a) Source: Plant perimeter values – SRS Annual Environmental Reports (see reference section).
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 167 of 188

C.2.2 Releases of Noble Gases

Argon-41 was released in large quantities during reactor operation. The potential dose 

resulting from submersion in this noble gas is external because the gas is not absorbed 

by the body. Calculated air concentrations were estimated for populated areas at SRS 

and then converted to organ and whole body doses. The annual doses (assuming 2,000 

hours per year of exposure) have been tabulated based on 50th percentile source term 

and presented in Tables C-20 and C-21. The geometric standard deviations provided in 

these tables were based on the estimated 95th percentile source terms using the 

relationship provided in Section 3.1.1. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 168 of 188

Table C-20 Argon-41 50th Percentile Dose(a) to Skin(b) and Bone Surfaces(c) Resulting 

From Reactor Releases (mrem per year) and geometric standard deviation

Year A 

Area GSD

C 

Area GSD

D

Area GSD

FArea GSD BG GSDH

Area GSD

S 

Area GSD

Z 

Area GSD

R 

Area GSD

1955 0.17 1.15 1.81 1.15 0.30 1.15 0.39 1.15 0.42 1.15 0.42 1.15 0.36 1.15 0.35 1.15 2.14 1.15

1956 0.25 1.15 4.08 1.15 0.48 1.15 0.67 1.15 0.68 1.15 0.64 1.15 0.54 1.15 0.51 1.15 2.61 1.15

1957 0.31 1.15 4.22 1.15 0.56 1.15 0.77 1.15 0.79 1.15 0.78 1.15 0.66 1.15 0.63 1.15 4.09 1.15

1958 0.36 1.15 5.04 1.15 0.66 1.15 0.91 1.15 0.94 1.15 0.92 1.15 0.78 1.15 0.75 1.15 4.90 1.15

1959 0.46 1.15 6.67 1.15 0.85 1.15 1.17 1.15 1.21 1.15 1.17 1.15 0.99 1.15 0.94 1.15 5.80 1.15

1960 0.45 1.15 6.61 1.15 0.81 1.15 1.15 1.15 1.19 1.15 1.18 1.15 1.00 1.15 0.96 1.15 7.12 1.15

1961 0.45 1.15 6.61 1.15 0.81 1.15 1.15 1.15 1.19 1.15 1.16 1.15 0.98 1.15 0.94 1.15 6.46 1.15

1962 0.48 1.15 7.33 1.15 0.85 1.15 1.23 1.15 1.27 1.15 1.24 1.15 1.05 1.15 1.00 1.15 7.18 1.15

1963 0.50 1.15 7.41 1.15 0.90 1.15 1.27 1.15 1.31 1.15 1.28 1.15 1.08 1.15 1.04 1.15 7.22 1.15

1964 0.42 1.15 6.50 1.15 0.75 1.15 1.09 1.15 1.13 1.15 1.11 1.15 0.94 1.15 0.90 1.15 7.04 1.15

1965 0.39 1.15 5.76 1.15 0.70 1.15 0.99 1.15 1.03 1.15 1.01 1.15 0.86 1.15 0.82 1.15 6.28 1.15

1966 0.31 1.15 5.62 1.15 0.63 1.15 0.84 1.15 0.84 1.15 0.74 1.15 0.61 1.15 0.56 1.15 0.72 1.15

1967 0.36 1.15 6.43 1.15 0.72 1.15 0.96 1.15 0.96 1.15 0.85 1.15 0.70 1.15 0.64 1.15 0.82 1.15

1968 0.26 1.15 6.11 1.15 0.53 1.15 0.77 1.15 0.74 1.15 0.62 1.15 0.51 1.15 0.46 1.15 0.57 1.15

1969 0.17 1.15 3.84 1.15 0.34 1.15 0.49 1.15 0.47 1.15 0.39 1.15 0.32 1.15 0.29 1.15 0.35 1.15

1970 0.13 1.15 2.83 1.15 0.26 1.15 0.37 1.15 0.36 1.15 0.31 1.15 0.25 1.15 0.23 1.15 0.29 1.15

1971 0.16 1.15 3.27 1.15 0.31 1.15 0.45 1.15 0.44 1.15 0.38 1.15 0.32 1.15 0.29 1.15 0.39 1.15

1972 0.19 1.15 4.11 1.15 0.40 1.15 0.55 1.15 0.53 1.15 0.46 1.15 0.38 1.15 0.34 1.15 0.44 1.15

1973 0.21 1.15 3.60 1.15 0.43 1.15 0.55 1.15 0.54 1.15 0.48 1.15 0.40 1.15 0.37 1.15 0.50 1.15

1974 0.12 1.15 2.05 1.15 0.23 1.15 0.31 1.15 0.31 1.15 0.28 1.15 0.23 1.15 0.22 1.15 0.32 1.15

1975 0.08 1.15 1.81 1.15 0.15 1.15 0.23 1.15 0.22 1.15 0.18 1.15 0.15 1.15 0.14 1.15 0.17 1.15

1976 0.10 1.15 2.43 1.15 0.20 1.15 0.30 1.15 0.29 1.15 0.24 1.15 0.19 1.15 0.18 1.15 0.22 1.15

1977 0.08 1.15 2.28 1.15 0.16 1.15 0.25 1.15 0.24 1.15 0.20 1.15 0.16 1.15 0.14 1.15 0.16 1.15

1978 0.06 1.15 1.33 1.15 0.13 1.15 0.18 1.15 0.17 1.15 0.15 1.15 0.12 1.15 0.11 1.15 0.14 1.15

1979 0.06 1.15 1.58 1.15 0.13 1.15 0.19 1.15 0.18 1.15 0.15 1.15 0.13 1.15 0.11 1.15 0.14 1.15

1980 0.09 1.15 2.35 1.15 0.16 1.15 0.27 1.15 0.25 1.15 0.21 1.15 0.17 1.15 0.15 1.15 0.18 1.15

1981 0.07 1.15 1.72 1.15 0.15 1.15 0.22 1.15 0.21 1.15 0.18 1.15 0.14 1.15 0.13 1.15 0.16 1.15

1982 0.07 1.15 1.50 1.15 0.14 1.15 0.20 1.15 0.19 1.15 0.17 1.15 0.14 1.15 0.12 1.15 0.17 1.15

1983 0.04 1.15 0.50 1.15 0.08 1.15 0.10 1.15 0.10 1.15 0.10 1.15 0.08 1.15 0.08 1.15 0.13 1.15

1984 0.04 1.15 0.54 1.15 0.07 1.15 0.09 1.15 0.10 1.15 0.09 1.15 0.08 1.15 0.07 1.15 0.12 1.15

1985 0.05 1.15 0.84 1.15 0.10 1.15 0.14 1.15 0.14 1.15 0.13 1.15 0.11 1.15 0.10 1.15 0.17 1.15

1986 0.06 1.15 0.19 1.15 0.12 1.15 0.13 1.15 0.14 1.15 0.15 1.15 0.13 1.15 0.13 1.15 0.24 1.15

1987 0.09 0.86 0.33 0.97 0.21 0.98 0.19 0.89 0.20 0.87 0.21 0.83 0.17 0.82 0.17 0.81 0.26 0.66

1988 0.03 0.79 0.11 0.91 0.07 0.91 0.07 0.82 0.08 0.80 0.08 0.76 0.07 0.75 0.06 0.74 0.11 0.60

1989 0.00 1.15 0.00 1.15 0.00 1.15 0.00 1.15 0.00 1.15 0.00 1.15 0.00 1.15 0.00 1.15 0.00 1.15

(a) Doses estimated based on 2,000 hours per year exposure and FGR #12 dose conversion factor of 1.01E-13 Sv m3

per sec Bq.

(b) Dose to skin should be assessed as 30% from >15 KeV electrons and 70% from >250 KeV photons.

(c) Dose to bone surfaces should be assumed to be 100% from >250 KeV photons. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 169 of 188

Table C-20 (cont.) Argon-41 50th Percentile Dose(a) to Skin(b) and Bone Surfaces(c) Resulting 

From Reactor Releases (mrem per year) and geometric standard deviation

Year

RR 

Yard GSD

P

Area GSD

L

Area GSD

KArea GSD FS GSD CS GSDAverage GSDMaximum GSD

1955 0.67 1.15 4.21 1.15 2.67 1.15 2.56 1.15 0.18 1.15 0.56 1.15 1.15 1.15 4.21 1.15

1956 0.78 1.15 4.09 1.15 4.27 1.15 4.23 1.15 0.26 1.15 1.02 1.15 1.67 1.15 4.27 1.15

1957 0.96 1.15 4.95 1.15 5.13 1.15 5.04 1.15 0.31 1.15 1.14 1.15 2.02 1.15 5.13 1.15

1958 1.13 1.15 5.78 1.15 5.99 1.15 5.90 1.15 0.37 1.15 1.35 1.15 2.39 1.15 5.99 1.15

1959 1.44 1.15 7.41 1.15 7.69 1.15 7.58 1.15 0.47 1.15 1.75 1.15 3.04 1.15 7.69 1.15

1960 1.38 1.15 6.72 1.15 6.93 1.15 6.84 1.15 0.47 1.15 1.72 1.15 2.97 1.15 7.12 1.15

1961 1.43 1.15 7.38 1.15 6.96 1.15 6.84 1.15 0.47 1.15 1.72 1.15 2.97 1.15 7.38 1.15

1962 1.48 1.15 7.44 1.15 7.01 1.15 6.91 1.15 0.49 1.15 1.85 1.15 3.12 1.15 7.44 1.15

1963 1.52 1.15 7.52 1.15 7.77 1.15 7.67 1.15 0.51 1.15 1.91 1.15 3.26 1.15 7.77 1.15

1964 1.26 1.15 5.95 1.15 6.12 1.15 6.05 1.15 0.44 1.15 1.64 1.15 2.76 1.15 7.04 1.15

1965 1.14 1.15 5.19 1.15 6.02 1.15 5.96 1.15 0.40 1.15 1.49 1.15 2.54 1.15 6.28 1.15

1966 0.94 1.15 5.56 1.15 5.88 1.15 5.84 1.15 0.29 1.15 1.34 1.15 2.05 1.15 5.88 1.15

1967 1.08 1.15 6.36 1.15 6.72 1.15 6.68 1.15 0.34 1.15 1.54 1.15 2.34 1.15 6.72 1.15

1968 0.77 1.15 5.23 1.15 1.01 1.15 5.49 1.15 0.24 1.15 1.28 1.15 1.64 1.15 6.11 1.15

1969 0.46 1.15 3.02 1.15 0.65 1.15 3.86 1.15 0.15 1.15 0.81 1.15 1.04 1.15 3.86 1.15

1970 0.40 1.15 2.73 1.15 0.51 1.15 2.85 1.15 0.12 1.15 0.61 1.15 0.82 1.15 2.85 1.15

1971 0.58 1.15 4.31 1.15 0.66 1.15 3.16 1.15 0.15 1.15 0.72 1.15 1.04 1.15 4.31 1.15

1972 0.60 1.15 4.18 1.15 0.79 1.15 4.45 1.15 0.18 1.15 0.90 1.15 1.23 1.15 4.45 1.15

1973 0.72 1.15 5.13 1.15 0.90 1.15 5.34 1.15 0.19 1.15 0.86 1.15 1.35 1.15 5.34 1.15

1974 0.49 1.15 3.77 1.15 0.52 1.15 2.36 1.15 0.11 1.15 0.49 1.15 0.79 1.15 3.77 1.15

1975 0.24 1.15 1.70 1.15 0.30 1.15 1.44 1.15 0.07 1.15 0.38 1.15 0.48 1.15 1.81 1.15

1976 0.29 1.15 1.99 1.15 0.37 1.15 1.90 1.15 0.09 1.15 0.50 1.15 0.62 1.15 2.43 1.15

1977 0.21 1.15 1.39 1.15 0.28 1.15 1.31 1.15 0.07 1.15 0.44 1.15 0.49 1.15 2.28 1.15

1978 0.19 1.15 1.30 1.15 0.25 1.15 1.42 1.15 0.06 1.15 0.29 1.15 0.39 1.15 1.42 1.15

1979 0.18 1.15 1.24 1.15 0.24 1.15 1.24 1.15 0.06 1.15 0.32 1.15 0.40 1.15 1.58 1.15

1980 0.24 1.15 1.67 1.15 0.30 1.15 1.31 1.15 0.08 1.15 0.46 1.15 0.53 1.15 2.35 1.15

1981 0.22 1.15 1.47 1.15 0.28 1.15 1.53 1.15 0.07 1.15 0.36 1.15 0.46 1.15 1.72 1.15

1982 0.24 1.15 1.76 1.15 0.28 1.15 1.33 1.15 0.07 1.15 0.32 1.15 0.45 1.15 1.76 1.15

1983 0.21 1.15 1.76 1.15 0.20 1.15 0.83 1.15 0.04 1.15 0.14 1.15 0.29 1.15 1.76 1.15

1984 0.20 1.15 1.67 1.15 0.17 1.15 0.50 1.15 0.04 1.15 0.14 1.15 0.26 1.15 1.67 1.15

1985 0.27 1.15 2.23 1.15 0.25 1.15 0.84 1.15 0.05 1.15 0.21 1.15 0.38 1.15 2.23 1.15

1986 0.44 1.15 3.76 1.15 0.36 1.15 1.27 1.15 0.07 1.15 0.15 1.15 0.49 1.15 3.76 1.15

1987 0.40 0.50 2.89 0.20 1.40 1.04 3.03 1.12 0.09 0.82 0.23 0.90 0.66 0.83 3.03 1.12

1988 0.18 0.45 1.31 0.18 0.71 1.05 0.80 1.11 0.03 0.75 0.09 0.83 0.25 0.75 1.31 0.82

1989 0.00 1.15 0.01 1.15 0.11 1.15 0.01 1.15 0.00 1.15 0.00 1.15 0.01 1.15 0.11 1.15

(a) Doses estimated based on 2,000 hours per year exposure and FGR #12 dose conversion factor of 1.01E-13 Sv m3

per sec Bq.

(b) Dose to skin should be assessed as 30% from >15 KeV electrons and 70% from >250 KeV photons.

(c) Dose to bone surfaces should be assumed to be 100% from >250 KeV photons.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 170 of 188

Table C-21 Argon-41 50th Percentile Dose(a) (b) to Whole Body and Other Organs 

Resulting From Reactor Releases (mrem per year) and geometric standard deviation

Year A

Area GSD

C

Area GSD

D

Area GSD

FArea GSD BG GSDH

Area GSD

S

Area GSD

Z

Area GSD

R 

Area GSD

1955 0.12 1.15 1.28 1.15 0.21 1.15 0.28 1.15 0.30 1.15 0.30 1.15 0.26 1.15 0.25 1.15 1.52 1.15

1956 0.18 1.15 2.89 1.15 0.34 1.15 0.47 1.15 0.48 1.15 0.45 1.15 0.38 1.15 0.36 1.15 1.85 1.15

1957 0.22 1.15 2.99 1.15 0.40 1.15 0.54 1.15 0.56 1.15 0.55 1.15 0.47 1.15 0.45 1.15 2.90 1.15

1958 0.26 1.15 3.58 1.15 0.47 1.15 0.64 1.15 0.67 1.15 0.65 1.15 0.55 1.15 0.53 1.15 3.47 1.15

1959 0.33 1.15 4.73 1.15 0.61 1.15 0.83 1.15 0.86 1.15 0.83 1.15 0.70 1.15 0.67 1.15 4.11 1.15

1960 0.32 1.15 4.68 1.15 0.58 1.15 0.82 1.15 0.85 1.15 0.84 1.15 0.71 1.15 0.68 1.15 5.05 1.15

1961 0.32 1.15 4.68 1.15 0.58 1.15 0.81 1.15 0.84 1.15 0.83 1.15 0.70 1.15 0.67 1.15 4.58 1.15

1962 0.34 1.15 5.20 1.15 0.60 1.15 0.87 1.15 0.90 1.15 0.88 1.15 0.74 1.15 0.71 1.15 5.09 1.15

1963 0.35 1.15 5.25 1.15 0.64 1.15 0.90 1.15 0.93 1.15 0.91 1.15 0.77 1.15 0.74 1.15 5.12 1.15

1964 0.30 1.15 4.61 1.15 0.53 1.15 0.77 1.15 0.80 1.15 0.79 1.15 0.67 1.15 0.64 1.15 4.99 1.15

1965 0.28 1.15 4.09 1.15 0.50 1.15 0.71 1.15 0.73 1.15 0.72 1.15 0.61 1.15 0.58 1.15 4.45 1.15

1966 0.22 1.15 3.99 1.15 0.45 1.15 0.60 1.15 0.59 1.15 0.52 1.15 0.43 1.15 0.40 1.15 0.51 1.15

1967 0.25 1.15 4.56 1.15 0.51 1.15 0.68 1.15 0.68 1.15 0.60 1.15 0.49 1.15 0.45 1.15 0.58 1.15

1968 0.19 1.15 4.33 1.15 0.37 1.15 0.54 1.15 0.52 1.15 0.44 1.15 0.36 1.15 0.33 1.15 0.40 1.15

1969 0.12 1.15 2.72 1.15 0.24 1.15 0.35 1.15 0.33 1.15 0.28 1.15 0.23 1.15 0.21 1.15 0.25 1.15

1970 0.09 1.15 2.00 1.15 0.19 1.15 0.26 1.15 0.25 1.15 0.22 1.15 0.18 1.15 0.16 1.15 0.20 1.15

1971 0.11 1.15 2.32 1.15 0.22 1.15 0.32 1.15 0.31 1.15 0.27 1.15 0.22 1.15 0.21 1.15 0.28 1.15

1972 0.14 1.15 2.91 1.15 0.28 1.15 0.39 1.15 0.38 1.15 0.32 1.15 0.27 1.15 0.24 1.15 0.31 1.15

1973 0.15 1.15 2.55 1.15 0.31 1.15 0.39 1.15 0.39 1.15 0.34 1.15 0.28 1.15 0.26 1.15 0.35 1.15

1974 0.08 1.15 1.45 1.15 0.16 1.15 0.22 1.15 0.22 1.15 0.20 1.15 0.17 1.15 0.15 1.15 0.22 1.15

1975 0.05 1.15 1.29 1.15 0.11 1.15 0.16 1.15 0.15 1.15 0.13 1.15 0.11 1.15 0.10 1.15 0.12 1.15

1976 0.07 1.15 1.73 1.15 0.14 1.15 0.21 1.15 0.20 1.15 0.17 1.15 0.14 1.15 0.12 1.15 0.15 1.15

1977 0.06 1.15 1.62 1.15 0.11 1.15 0.18 1.15 0.17 1.15 0.14 1.15 0.11 1.15 0.10 1.15 0.12 1.15

1978 0.04 1.15 0.95 1.15 0.09 1.15 0.13 1.15 0.12 1.15 0.10 1.15 0.09 1.15 0.08 1.15 0.10 1.15

1979 0.05 1.15 1.12 1.15 0.09 1.15 0.14 1.15 0.13 1.15 0.11 1.15 0.09 1.15 0.08 1.15 0.10 1.15

1980 0.06 1.15 1.66 1.15 0.12 1.15 0.19 1.15 0.18 1.15 0.15 1.15 0.12 1.15 0.11 1.15 0.13 1.15

1981 0.05 1.15 1.22 1.15 0.10 1.15 0.15 1.15 0.15 1.15 0.12 1.15 0.10 1.15 0.09 1.15 0.11 1.15

1982 0.05 1.15 1.06 1.15 0.10 1.15 0.14 1.15 0.14 1.15 0.12 1.15 0.10 1.15 0.09 1.15 0.12 1.15

1983 0.03 1.15 0.35 1.15 0.06 1.15 0.07 1.15 0.07 1.15 0.07 1.15 0.06 1.15 0.06 1.15 0.09 1.15

1984 0.03 1.15 0.39 1.15 0.05 1.15 0.07 1.15 0.07 1.15 0.06 1.15 0.05 1.15 0.05 1.15 0.08 1.15

1985 0.04 1.15 0.60 1.15 0.07 1.15 0.10 1.15 0.10 1.15 0.09 1.15 0.08 1.15 0.07 1.15 0.12 1.15

1986 0.04 1.15 0.14 1.15 0.08 1.15 0.09 1.15 0.10 1.15 0.11 1.15 0.09 1.15 0.09 1.15 0.17 1.15

1987 0.06 0.86 0.24 0.97 0.15 0.98 0.13 0.89 0.14 0.87 0.15 0.83 0.12 0.82 0.12 0.81 0.19 0.66

1988 0.02 0.79 0.08 0.91 0.05 0.91 0.05 0.82 0.05 0.80 0.06 0.76 0.05 0.75 0.05 0.74 0.08 0.60

1989 0.00 1.15 0.00 1.15 0.00 1.15 0.00 1.15 0.00 1.15 0.00 1.15 0.00 1.15 0.00 1.15 0.00 1.15

(a) Doses estimated based on 2,000 hours per year exposure and FGR #12 dose conversion factor of 7.16E-14 Sv m3

per sec Bq.

(b) Dose to whole body and other underlying organs should be assumed to be 100% from >250 KeV photons.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 171 of 188

Table C-21 (cont.) Argon-41 95th Percentile Dose (a) (b) to Whole Body and Other Organs 

Resulting From Reactor Releases (mrem per year) and geometric standard deviation

Year

RR 

Yard GSD

P 

Area GSD

L 

Area GSD

KArea GSD FS GSD CS GSDAverage GSDMaximum GSD

1955 0.47 1.15 2.99 1.15 1.90 1.15 1.81 1.15 0.12 1.15 0.40 1.15 0.81 1.15 2.99 1.15

1956 0.55 1.15 2.90 1.15 3.03 1.15 3.00 1.15 0.18 1.15 0.72 1.15 1.19 1.15 3.03 1.15

1957 0.68 1.15 3.51 1.15 3.63 1.15 3.58 1.15 0.22 1.15 0.81 1.15 1.43 1.15 3.63 1.15

1958 0.80 1.15 4.10 1.15 4.25 1.15 4.18 1.15 0.26 1.15 0.96 1.15 1.69 1.15 4.25 1.15

1959 1.02 1.15 5.26 1.15 5.45 1.15 5.38 1.15 0.33 1.15 1.24 1.15 2.16 1.15 5.45 1.15

1960 0.98 1.15 4.76 1.15 4.91 1.15 4.85 1.15 0.33 1.15 1.22 1.15 2.10 1.15 5.05 1.15

1961 1.01 1.15 5.23 1.15 4.93 1.15 4.85 1.15 0.33 1.15 1.22 1.15 2.11 1.15 5.23 1.15

1962 1.05 1.15 5.28 1.15 4.97 1.15 4.90 1.15 0.35 1.15 1.31 1.15 2.21 1.15 5.28 1.15

1963 1.08 1.15 5.33 1.15 5.51 1.15 5.44 1.15 0.36 1.15 1.35 1.15 2.31 1.15 5.51 1.15

1964 0.90 1.15 4.22 1.15 4.34 1.15 4.29 1.15 0.31 1.15 1.16 1.15 1.96 1.15 4.99 1.15

1965 0.81 1.15 3.68 1.15 4.27 1.15 4.23 1.15 0.29 1.15 1.05 1.15 1.80 1.15 4.45 1.15

1966 0.67 1.15 3.94 1.15 4.17 1.15 4.14 1.15 0.21 1.15 0.95 1.15 1.45 1.15 4.17 1.15

1967 0.76 1.15 4.51 1.15 4.76 1.15 4.73 1.15 0.24 1.15 1.09 1.15 1.66 1.15 4.76 1.15

1968 0.54 1.15 3.71 1.15 0.72 1.15 3.89 1.15 0.17 1.15 0.91 1.15 1.16 1.15 4.33 1.15

1969 0.33 1.15 2.14 1.15 0.46 1.15 2.73 1.15 0.11 1.15 0.57 1.15 0.74 1.15 2.73 1.15

1970 0.28 1.15 1.94 1.15 0.36 1.15 2.02 1.15 0.08 1.15 0.43 1.15 0.58 1.15 2.02 1.15

1971 0.41 1.15 3.06 1.15 0.47 1.15 2.24 1.15 0.11 1.15 0.51 1.15 0.74 1.15 3.06 1.15

1972 0.43 1.15 2.96 1.15 0.56 1.15 3.16 1.15 0.13 1.15 0.64 1.15 0.87 1.15 3.16 1.15

1973 0.51 1.15 3.64 1.15 0.64 1.15 3.79 1.15 0.14 1.15 0.61 1.15 0.96 1.15 3.79 1.15

1974 0.35 1.15 2.67 1.15 0.37 1.15 1.67 1.15 0.08 1.15 0.35 1.15 0.56 1.15 2.67 1.15

1975 0.17 1.15 1.20 1.15 0.21 1.15 1.02 1.15 0.05 1.15 0.27 1.15 0.34 1.15 1.29 1.15

1976 0.21 1.15 1.41 1.15 0.27 1.15 1.35 1.15 0.06 1.15 0.35 1.15 0.44 1.15 1.73 1.15

1977 0.15 1.15 0.98 1.15 0.20 1.15 0.93 1.15 0.05 1.15 0.31 1.15 0.35 1.15 1.62 1.15

1978 0.13 1.15 0.92 1.15 0.18 1.15 1.00 1.15 0.04 1.15 0.20 1.15 0.28 1.15 1.00 1.15

1979 0.13 1.15 0.88 1.15 0.17 1.15 0.88 1.15 0.04 1.15 0.23 1.15 0.28 1.15 1.12 1.15

1980 0.17 1.15 1.19 1.15 0.21 1.15 0.93 1.15 0.06 1.15 0.33 1.15 0.37 1.15 1.66 1.15

1981 0.15 1.15 1.04 1.15 0.20 1.15 1.09 1.15 0.05 1.15 0.26 1.15 0.33 1.15 1.22 1.15

1982 0.17 1.15 1.25 1.15 0.20 1.15 0.94 1.15 0.05 1.15 0.23 1.15 0.32 1.15 1.25 1.15

1983 0.15 1.15 1.24 1.15 0.14 1.15 0.59 1.15 0.03 1.15 0.10 1.15 0.21 1.15 1.24 1.15

1984 0.14 1.15 1.18 1.15 0.12 1.15 0.35 1.15 0.03 1.15 0.10 1.15 0.18 1.15 1.18 1.15

1985 0.19 1.15 1.58 1.15 0.18 1.15 0.60 1.15 0.04 1.15 0.15 1.15 0.27 1.15 1.58 1.15

1986 0.31 1.15 2.66 1.15 0.26 1.15 0.90 1.15 0.05 1.15 0.11 1.15 0.35 1.15 2.66 1.15

1987 0.29 0.50 2.05 0.20 0.99 1.04 2.15 1.12 0.06 0.82 0.16 0.90 0.47 0.83 2.15 1.12

1988 0.12 0.45 0.93 0.18 0.50 1.05 0.57 1.11 0.02 0.75 0.06 0.83 0.18 0.75 0.93 0.82

1989 0.00 1.15 0.01 1.15 0.08 1.15 0.01 1.15 0.00 1.15 0.00 1.15 0.01 1.15 0.08 1.15

(a) Doses estimated based on 2,000 hours per year exposure and FGR #12 dose conversion factor of 7.16E-14 Sv m3 per sec 

Bq. 

(b) Dose to whole body and other underlying organs should be assumed to be 100% from >250 KeV photons.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 172 of 188

ATTACHMENT D 

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

D.0 Occupational Internal Dose

The bioassay program monitored internal uptake of radionuclides using both in vitro

analysis of urine and feces and in vivo monitoring using whole body and chest counting.

D.1 In vitro Minimum Detectable Activities (MDAs) and Reporting Levels

The decision level is the level at which activity is considered present in a sample with a

95% confidence level. The reporting level is the minimally acceptable decision level. It 

appears from the database that at some times and certain radionuclides the reporting 

level was also the level below which a result was reported as “less than”, regardless of 

the actual decision level for a given sample. Since 1990, the MDA has been defined as 

the quantity of material (or activity) that has a 5% chance of not being detected. Prior to 

1990, the term MDA may have had less rigorous or at least slightly different meanings. 

For some times and certain radionuclides, MDA and reporting level were synonymous. 

For the dosimetrist the reporting level is the relevant concept over the history of the SRS 

records. There is no consistent quantitative relationship between the reporting level and 

the MDA.

D.1.1 In vitro Urine Analysis

The reporting levels and MDAs (if known and different than reporting levels), are shown 

in Table D-01. These values are applicable over the time period specified. The lack of an 

MDA in a given time period reflects the lack of documented information about urine 

analysis for that radionuclide. Activity fractions for depleted, recycled, and high enriched 

uranium are listed in the glossary section. Activity fractions for 6% and 12% plutonium

mixtures are given in Chapter 4.

The following codes are found in the excreta records:

C routine urine sample, includes baselines and termination samples

A-A 24-hours special urine sample

A-O two 24-hour urine samples, one analyzed, one held 

A-W follow-up to positive routine sample

IA Inconclusive Analysis; analysis didn’t meet QC criteria; no result reported

B follow-up to an IA sample

A-F single void fecal sample

T tritium sample, only 50 mL collected

P-10 an old code for tritium

SP Neptunium

FP fission products; gamma scan.

FPIA fission products/induced activity. Same analysis as FP.

IA with a numerical result would be same as FP

LMF uranium
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 173 of 188

D.1.2 In vitro Urine Analyses for Individual Radionuclides

Urine analysis for trivalent actinides (americium, curium, and californium)

Urinalysis for trivalent actinides began in the mid-1960s. From initiation to 1990, the 

reporting levels varied from about 1 dpm/1.5L to 0.3 dpm/L. In 1990, the reporting levels 

decreased to 0.1 dpm/L. The MDAs since 1994 are shown in Table D-01. The counting 

method does not distinguish 241Am from 244Cm from 252Cf. Interpretation of which 

radionuclide is being counted will have to come from particulars of the intake description. 

The claimant-favorable default is 241Am.

Urine analysis for plutonium

The reporting levels and MDAs are also listed in Table D-01. Results prior to 1981 (1988 

for special samples) were for total alpha from plutonium isotopes. Plutonium-239 results 

are usually indicative of a mixture of plutonium isotopes. See 4.1.2 for discussion. 

Urine analysis for tritium

From startup until 1958, the reported MDA was 1 μCi/L. From 1958 until present, the 

MDA improved to the current level of 20 nCi/L; however, the reporting level remained at 

1 μCi/ L, and decreased to the current value of 0.1 μCi/L. For tritium, the denominator 

has always been per liter of urine, and not 1.5 L as for other radionuclides.

For current analyses, tritium results of ≤0.05 μCi/L are reported as <0.1μCi/L, from 0.05 

to 0.1 μCi/L are reported as 0.1 μCi/L, and >0.1 μCi/ L are reported as measured. Only 

results greater than 5 μCi/L were evaluated prior to computer evaluation of data.

Urine analysis for uranium

A variety of methods has been used historically to analyze uranium resulting in 

differences in the MDA over time, as shown in Table D-01. Activity fractions for depleted, 

recycled, and highly enriched uranium are listed in the glossary section. For earlier 

monitoring periods, the designations “enriched” and “depleted” analysis for uranium 

referred to the analysis and were not indicative of the degree of uranium enrichment.

Natural background levels of uranium in excreta samples can interfere with interpretation 

of bioassay results; see the discussion in 4.1.2. 

Urine analysis for radiostrontium

The start date for urinalysis for radioisotopes of strontium was not specifically 

determined, other than “late 1950s.” This was referred to as the fission product analysis. 

From start until 1969 the same method was used, which involved separation of the 

strontium and beta counting. Both 89Sr and 90Sr would have been counted, but it is 

claimant-favorable to assume the result is all 90Sr. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 174 of 188

D.1.3 In vitro Fecal Analysis

Fecal analysis has been done periodically since the 1950s for special purposes and was 

always done in conjunction with urine bioassay sampling. Usually the analysis was a 

gamma scan on the total sample with typical MDAs of 7 pCi for 241Am, 300 pCi for 238Pu, 

and 600 pCi for 238Pu plus 239Pu in weapons grade mixtures. If fecal samples were also 

analyzed by wet radiochemistry, the MDAs were approximately 1.5 times the same 

MDAs for urinalyses.

Table D-01 Limits of Detection for Urinalysisa 

Radionuclide Time Period MDA (pCi/L) Reporting Level (pCi/L)

3

H Start to 1958 1 E+06

3

H 1958 to 2002 5 E+05

3

H Present 2.0E+04 1E+05b

Sr Start to 1969 (see fission product)

90Sr Present 2.9 4

FP 1955 to 1960 30 dpm/75mL

FP 1960-1965 100 dpm/1.5L (beta)

500 dpm/1.5L (gamma)

FP 1965 - ? 1 nCi/1.5L

234U 1990 to present 0.032

235U 1982 to 1986 0.14 ng NA

235U 1990 to present 0.036

238U 1990 to present 0.032

U-natural 1982 to 1986 1 μg/L 1 μg/L

DU 1954 to 1982 Not listed Not listed

DU 1982 to 1986 same as U-nat same as U-nat

DU 1986-1994 0.5/1.5L 1 dpm/1.5L 

RU 1990 to present use 234U and 

isotopic ratios

use 234U and isotopic ratios

EU Start to mid-1960s 0.15 dpm/1.5 L

EU mid-1960s to 1982 1 dpm/1.5 L

EU 1982 to 1986 1 dpm/L

EU 1986 to 1990 1dpm/1.5 L

HEU 1990 to present use 235U and 

isotopic ratios

use 235U and isotopic ratios

SP (237Np) ~1959 to 1964 0.05 dpm/1.5L

SP 1964 to 2001 0.1 dpm/1.5L

237Np 2001 to present 0.035 c

238,239,240Pu 1954 to 1962 0.05dpm/1.5L

238,239,240Pu 1963 to 1981 0.1dpm/1.5 L

238Pu 1981 to 1988d 0.05/1.5 L

238

Pu 1988 to present 0.07

239Pu, 240Pu 1981 to 1988d 0.07/1.5 L
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 175 of 188

Radionuclide Time Period MDA (pCi/L) Reporting Level (pCi/L)

239Pu, 240Pu 1988 to present 0.06

241Am Mid 1960s to1971 3 dpm/1.5 L

241Am 1971 to 1990 0.3 dpm/1.5 L

241Am 1990 to 1994 0.1 dpm/L

241Am 1994 to present 0.029 

244Cm Mid 1960s to1971 3 dpm/1.5 L

244Cm 1971 to 1990 0.3 dpm/1.5 L

244Cm 1990 to 1994 0.1 dpm/L

244Cm 1994 to present 0.021

252Cf Mid 1960s to1971 3 dpm/1.5 L

252Cf 1971 to 1990 0.3 dpm/1.5 L

252Cf 1990 to 1994 0.1 dpm/L

252Cf 1994 to present 0.021

(a) MDAs are given if less than reporting levels; generally the reporting level should be used as the limit of 

detection.

(b) Measured values between 5E+4 and 1E+5 pCi/L are reported in the database as 0.1 μCi/L.

(c) Some uncertainty about this value. See discussion in Chapter 4.

(d) For routine samples; gross alpha method still used for special samples. 

D.2 In vivo MDAs and reporting practices at SRS

Whole body counting has been performed since 1960 and chest counting was initiated in 

the early 1970s. In vivo detection capabilities are reviewed below.

D.2.1 Whole Body Counting

A variety of hardware and software applications have been used for whole body counting 

since 1960 consistent with industry-wide improvements in the discipline. Whole body 

counting was not used for plutonium and americium bioassay due to their low energy 

photon emissions. Sometimes MDAs were reported on the forms for the individual 

measurements and these should be used when available. Default MDAs for various 

radionuclides are shown in Table D-02. MDAs and reporting levels were synonymous. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 176 of 188

Table D-02 MDAs for Whole Body Counting

Radionuclide Time Period MDA (nCi)

54Mn Present 3.4

60Co Present 2.9

65Zn 1960 to ?? 5.1

65Zn Present 6.1

Zr/95Nb 1960 to ?? 2.2

106Ru 1960 to ?? 6.1

106Ru Present 36

125Sb Present 14

131I 1960 to ?? 1.4

134Cs Present 3.8

137Cs 1960 to ?? 1.0

137Cs Present 4.1

Ba/140La 1960 to ?? 9.3

144Ce 1960 to ?? 29

144Ce Present 69

152Eu Present 18

154Eu Present 8.4

235U Present 14

U-natural (235U) 1960 to ?? 62

237Np Present 14

For special investigational counts, the MDAs are approximately 0.71 times these values 

for whole body counting

D.2.2 Chest Counting

No MDA information could be found for chest counting performed from 1966 to around 

1971. From 1971 to 1987 or 1988, using phoswich detectors, chest burdens of 4 to 10 

nCi were reported for 238Pu. In 1987 or 1988, the 238Pu MDA is reported to be 60-70 nCi.

MDAs for later periods have been generated individually for counts by processing 

software, with typical values shown in Table D-03.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 177 of 188

Table D-03 MDAs for Chest Counting

Radionuclide Time Period MDA (nCi)

144Ce Present 0.31

152Eu Present 0.056

228Th Present 3.2

232Th Present 31

234U Present 30

235U Present 0.10

236U Present 89

238U Present 1.1

DU Present 1.2

RU Present 8.3

HEU Present 5.2

237Np Present 0.31

238Pu <1971 NA

238Pu 1971 to 1987 or 1988 4 to 10

238Pu 1987 to Present 130a 

239Pu 1987 to Present 290a 

Pu240 1987 to Present 110a 

241Am 1987 to Present 0.13a

243Am 1987 to Present 0.12

244Cm 1987 to Present 78a 

252Cf 1987 to Present 85a 

Based on chest wall thickness of 50th male worker of 3.7 cm

For special investigational counts, the MDAs are approximately 0.77 times the values for chest counting in 

Table D-03.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 178 of 188

ATTACHMENT E 

OCCUPATIONAL EXTERNAL DOSE FOR MONITORED WORKERS

E.0 Occupational External Dose

The information needed to evaluate claims is directed to the technical parameters of the 

annual estimates of the primary organ dose that is calculated from the dosimeter 

interpreted personal dose equivalent, Hp(10), and, in the case of skin, testicular and 

breast cancer, Hp(0.07), used as a consistent basis of comparison for all years of SRS 

occupational radiation exposure. 

The primary IREP screen used to input dose parameters is illustrated in Table E-01. 

Input to these fields is obtained from the SRS dose of record. The claim provides the 

primary organ of interest and other worker information needed to run IREP. Guidance to 

the dose reconstruction analyst to select technical external dose parameters to complete 

the respective fields in Table E-01 is presented in the following sections. 

Table E-01 IREP Dose Parameter Input Screen

Exposure Radiation TypeDistribution Parameters

# Year Rate Type 1 2 3

1 1960 Acute Photon, 30-250 keV Normal 2 2 0

2 1961 Acute

E.1 Years of Exposure

The years of exposure should be identified from the SRS radiation dose reports. Missed 

dose is calculated for all zero or missing records within a reporting year as noted by nonzero or blank entries. SRS policies require monitoring of all workers occupationally 

exposed. Missed dose should be calculated for a claimant for each year of record as an 

employee unless there are valid reasons for years in which there is no SRS recorded 

dose. 

E.2 Rate 

As identified in OCAS-IG-001, acute is selected for all types of external beta and photon 

dose. Chronic is selected for tritium and neutron dose.

E.3 Radiation Type

E.3.1 Beta/Photon

Claimant-favorable assumptions should be made using guidance in Table E-02 for beta 

and photon (i.e., x-ray and gamma rays) radiation types to assure that dose is not 

underestimated. The values presented in this table are intended to provide a reasonable 

estimate of parameters used to calculate the organ dose without significant numerical 

error for long-term SRS workers in the respective facilities. There is no direct evidence to 

select the specific values shown other than considerations of the radiation sources and 

usual work tasks. In those cases where there is some doubt in the values, a range of 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 179 of 188

realistic values should be selected for comparison and the most claimant-favorable 

option selected.

Table E-02 Selection of Beta and Photon Radiation Energies and Percentages for SRS 

Facilities and Processes

Process/

Buildings Description

Operations Radiation 

Type

Energy 

Begin End Selection Percent

Fuel 

Fabrication

Produced reactor fuel and target 

assemblies. 1953 1989

Beta

Photon

> 15 keV

30 – 250 keV

100%

100%

300M

Uranium fuel fabrication 1953 1989

Reactors

During Operation: Highly dispersed fields of higher-energy 

photon radiation fields from fission process, activation/ fission 

product nuclides. Potential for significant airborne nuclides and 

there may be significant higher-energy beta radiation. 

Beta

Photon

> 15 keV

30 – 250 keV

> 250 keV

100%

50%

50%

While not in Operation: Highly dispersed fields of higher-energy 

photon radiation fields from activation/fission product nuclides. 

There may be significant higher-energy beta radiation during 

maintenance work resulting from fission products.

105C

Heavy Water Moderated Reactors

3/1955 6/1985

105K 10/1954 1988

1992 7/1992

105L 7/1954 1968

1985 6/1988

105P 2/1954 8/1988

105R 12/1953 6/1964

Processing 

Plants

Radiochemical Operations: Highly dispersed fields of higherenergy photon radiation fields from activation/fission product 

nuclides dominant to most exposure profiles. Potential for higherenergy beta radiation during sampling and, maintenance work 

resulting from fission products. 

Beta

Photon

> 15 keV

30 – 250 keV

> 250 keV

100%

50%

50% 200F Radiochemical processing plant 1954

200H Radiochemical processing plant 1955

Plutonium 

Production

Plutonium Component Production: Plutonium is machined into 

weapon components using a glove box assembly process with 

predominant close anterior exposure to workers. Radiation 

characteristics in this area involve significant lower-energy 

photons and lower-energy neutron radiation.

Photon < 30 keV

30 – 250 keV

25%

75%

Plutonium Storage: Radiation characteristics in this area 

generally involve dispersed lower-energy neutron radiation and 

scattered photons to include 60 keV Am-241 gamma ray.

221F, B-line Plutonium Finishing Process 1954

221H, B-line Plutonium Finishing Process 1955

772F Production Control Laboratory 1955

235F Plutonium Fuel Facility 1955

736A Calibration Facility 1953

Radionuclide

Production

Radiation characteristics are highly dependent upon the 

radionuclide being produced. SRS has produced several 

radionuclides including tritium, 238Pu, 239Pu, americium, curium, 

californium, etc. 

(Specific to 

radio-nuclide)

Photon, 

default

30 – 250 keV

> 250 keV

 

50%

50%

773A Savannah River Laboratory 1953

Waste 

Handling

Radiation characteristics are highly dependent upon the source of 

waste but typically fission product nuclides (90Sr/90Y, 137Cs) are 

dominant. 

Beta

Photon

> 15 keV

30 – 250 keV

> 250 keV

100%

50%

50%

200F General waste handling, storage 

tanks, etc.

1953 Ongoing

200H General waste handling, storage 

tanks, etc

1953 Ongoing
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 180 of 188

E.3.2 Neutron

Table E-03 summarizes the default neutron dose distributions by energy for each 

neutron exposure area. 

Table E-03 Selection of Neutron Radiation Energies and Percentages for SRS Facilities 

and Processes

Process Description / Buildings Operations Neutron 

Energy

Default Dose 

Begin End Fraction (%)

Fuel 

Fabrication

Produced reactor fuel and target assemblies. Intermittent production mixed oxide fuel and 

plutonium – aluminum (Pu-Al) alloy targets.

(321-M)

0.1-2 MeV

2-20 MeV

83%

17%

Reactors

During Reactor Operation: Low-level neutron exposure through shielding. In addition, 

some intermittent neutron exposure from mixed oxide fuel and targets. 

(105C) 3/1955 1/1987

10-100 keV

0.1 – 2 MeV

15%

85%

(105K)

10/1954 1/1988

1992 7/1992

(105L)

7/1954 2/1968

1985 6/1988

(105P) 2/1954 8/1988

(105R) 12/1953 6/1964

Plutonium 

Production

Plutonium Finishing Process: Plutonium enters the process as PuF4 and is then fired into 

production pucks. Work is primarily conducted in glove boxes with predominant close 

anterior exposure to workers. Radiation levels at the beginning of the process are fairly 

constant while levels at end of process are closely related to production levels. 

Plutonium Finishing Process

(221-H, B-Line)

11/1954 1988

0.1-2 MeV

2-20 MeV

60%

40%

Plutonium Finishing Process

(221-F, B-Line) 10/1955 1988 0.1-2 MeV 100%

Plutonium Storage

0.1-2 MeV

2-20 MeV

86%

14%

Production Control Laboratory

(772-F) 1955 0.1-2 meV 100%

Plutonium Fuel Facility

(235-F)

1955 0.1-2 meV 100%

Radionuclid

e 

Production 

and 

Calibration

Radiation characteristics are highly dependent upon the radionuclide being produced. SRS 

has produced several radionuclides including 238Pu, 239Pu, americium, curium, californium, 

etc. At the calibration facility radiological instruments are calibrated using a variety of 

sources with different neutron energy spectra. 

 

Californium Production Facility 

(773-A)

0.1-2 MeV

2-20 MeV

50%

50%

Calibration Facility (736-A) 0.1-2 MeV

2-20 MeV

83%

17%
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 181 of 188

E.4 Distribution Parameters

Selection of the distribution parameters in Table E-01 involves adjustments to the dose 

of record for missed dose prior to entry into the IREP input screen. The selection of a 

normal distribution for the type determines the definition of Parameters #1 and #2. For a 

normal distribution, Parameter #3 is not used.

E.4.1 Parameter #1

For a normal distribution, parameter #1 is the mean of the distribution of recorded dose 

for each year of monitoring. Prior to calculating this, it is necessary to adjust the 

recorded dose for exposure profiles described in Section E.4.1.1 where the recorded 

dose of record will under-estimate Hp(10). 

E.4.1.1 Resolution of Photon, Neutron and Tritium Dose

For workers who terminated SRS employment prior to 1979, SRS hard-copy radiation 

records must be consulted to separate the photon, neutron and tritium dose for workers. 

Hard copy records do separate the recorded whole body dose (i.e., photon + neutron + 

tritium) with the exception of the period from the second quarter of 1963 through 

December 31, 1972. The missing dose components of the total whole body dose could 

be estimated from the average of the prior year (i.e., before second quarter of 1963) and 

subsequent year (after 1972) dose results according to the proportion of dose from 

photon, neutron and tritium dose based on examination of continuity in the worker’s job 

and work activities. A claimant-favorable option is simply to assign the whole body dose 

(i.e., photon + neutron + tritium) as a photon dose, and using the methods in the 

following sections a claimant-favorable neutron dose will also be calculated assuming 

the worker was working in a facility with neutron exposure.

E.4.1.2 Adjustments to Recorded Photon Dose

Adjustments to the SRS reported photon dose are necessary to arrive at a claimantfavorable dose considering the uncertainty associated primarily with the complex 

workplace radiation fields and exposure orientations. Taylor et al (1995) identified 

calibration corrections of 1.119 and 1.039 to the reported SRS TLD-Deep photon dose 

for the periods prior to 1986 and 1987, respectively, to estimate the Hp(10) dose. This 

is summarized in Table E-04.

Table E-04 Adjustments to Reported SRS Deep Photon Dose

Time 

Period Dosimeter Facility Step Adjustment to Reported Dose

Prior to 

1987

All 

beta/photon 

dosimeters

All facilities A Multiply reported TLD-Deep photon 

dose by a factor of 1.119 to estimate 

Hp(10).

For the year 

of 1987

TLD 

beta/photon 

dosimeter

All facilities B Multiply reported TLD-Deep photon 

dose by a factor of 1.039 to estimate 

Hp(10).

Post 1987 TLD 

beta/photon 

dosimeter

All facilities No adjustments made. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 182 of 188

E.4.1.3 Adjustments to Recorded Neutron Dose

The Savannah River Site (SRS) incorporated the energy variation of the dose equivalent 

into their calibration methodology. As a result, the recorded neutron dose is a 

combination of all neutron energies. In order to calculate the dose input to IREP (Table 

E-01), the recorded neutron dose must be separated into neutron energy groups as 

shown in Table E-03 and subsequently converted to ICRP 60 (1990) methodology. 

 

The neutron dose equivalent correction factor can be calculated by dividing the dose 

fractions from Table E-04 for each facility and neutron energy category group shown in 

Table E-02 by multiplying the recorded neutron dose by the area specific correction 

factors the neutron dose equivalent is calculated. For example consider a 1000 mrem 

recorded neutron dose by an energy employee working at the HB-Line, the corrected 

neutron dose is 1144 mrem from neutrons between 0.1-2.0 MeV and 529 mrem from 

neutrons with energy between 2 and 20 MeV. Thus the corrected neutron dose is a total 

of 1673 mrem. These corrections should be applied to measured dose, missed dose 

and dose determined based on a neutron to photon ratio. Table E-05 summarizes the 

dose fractions and associated correction factors for each neutron exposure area. 

E.4.1.4 Unmonitored Photon Dose

Adjustments to the recorded annual dose can be made using dose results for co-workers 

or using the recorded dose prior to and after the period of missed dose. These 

situations do require careful examination. 

E.4.1.5 Missed Photon Dose

The missed photon dose for dosimeter results less-than the MDL is particularly important 

for earlier years when MDLs were higher and dosimeter exchange was more frequent. 

OCAS-IG-001 guidance should be followed to calculate the missed dose by using a 

claimant favorable maximum potential missed dose. This is calculated using the 

Minimum Detection Level (MDL) multiplied by the number of zero dose results to provide 

an estimate of the maximum potential missed dose. The following sections describe the 

potential missed photon dose corrections according to facility/location, dosimeter type, 

year and energy range.
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 183 of 188

Table E-05 Summary of neutron Dose Fractions and associated ICRP 60 correction 

factors

Process Description / Buildings

Operations

Neutron 

Energy

Default 

Dose 

Fraction 

(%)

ICRP 60 

Correction 

Factor Begin End

Fuel 

Fabrication

Produced reactor fuel and target assemblies. Intermittent production mixed oxide fuel and plutonium – 

aluminum (Pu-Al) alloy targets.

(321-M)

0.1-2 MeV

2-20 MeV

83%

17%

1.58

0.23

Reactors

During Reactor Operation: Low-level neutron exposure through shielding. In addition, some intermittent 

neutron exposure from mixed oxide fuel and targets. 

(105C) 3/1955 1/1987

10-100 keV

0.1 – 2 MeV

15%

85%

0.28

1.62

(105K)

10/1954 1/1988

1992 7/1992

(105L)

7/1954 2/1968

1985 6/1988

(105P) 2/1954 8/1988

(105R) 12/1953 6/1964

Plutonium 

Production

Plutonium Finishing Process: Plutonium enters the process as PuF4 and is then fired into production 

pucks. Work is primarily conducted in glove boxes with predominant close anterior exposure to workers. 

Radiation levels at the beginning of the process are fairly constant while levels at end of process are 

closely related to production levels. 

Plutonium Finishing Process

(221-H, B-Line)

11/1954 1988

0.1-2 MeV

2-20 MeV

60%

40%

1.14

0.53

Plutonium Finishing Process

(221-F, B-Line)

10/1955 1988 0.1-2 MeV 100% 1.91

Plutonium Storage

0.1-2 MeV

2-20 MeV

86%

14%

1.64

0.19

Production Control Laboratory

(772-F)

1955

Plutonium Fuel Facility

(235-F)

1955

Radionuclid

e 

Production 

and 

Calibration

Radiation characteristics are highly dependent upon the radionuclide being produced. SRS has produced 

several radionuclides including 238Pu, 239Pu, americium, curium, californium, etc. At the calibration facility 

radiological instruments are calibrated using a variety of sources with different neutron energy spectra. 

 

Californium Production Facility 

(773-A)

0.1-2 MeV

2-20 MeV

50%

50%

Calibration Facility (736-A) 0.1-2 MeV

2-20 MeV

83%

17%

E.4.1.5.1 Facility/Location

The potential missed photon dose for the laboratory determined MDL and exchange 

frequency is presented in Table E-06. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 184 of 188

Table E-06 Missed photon dose according to SRS Facility.

SRS Facility 

Type Dosimeter Period of Use

Exchange 

Frequency

MDL 

(mSv)(a)

Max. Annual 

Missed Dose

(mSv)(b)

Fuel 

Fabrication, 

SRS Reactors, 

Processing 

Plants, Waste 

Handling

ORNL TwoElement film 

Dosimeter

Prior to 3/52 Weekly 0.4 20.8 

SRS Two 

Element film 

Dosimeter

After 3/52 through 

9/30/1957

Weekly 0.4 20.8 

10/1/1957 through 

11/8/1959

Biweekly 0.4 10.4 

Above plus 

plutonium and 

nuclide 

production 

SRS MultiElement Film 

Dosimeter

11/9/1959 through 

12/31/1964

Biweekly 0.4 10.4

1/1/1965 through 

12/31/1965

4-week 0.4 5.2

1/1/1966 through 

3/31/1970

Monthly 0.4 4.8

All SRS SRS TLD 4/1/1970 through 

6/30/1983

Monthly 0.15 1.8

Panasonic TLD 7/1/1983 to 2003 

(ongoing)

Monthly 0.05 0.6

Panasonic TLND 1/1/1995 to 2003 

(ongoing)

Monthly 

(n=12)

0.15 0.9

a. Estimated film dosimeter detection levels based on NIOSH (1993), NRC (1989), and Wilson et al (1990). 

TLD detection levels from Taylor et al (1995) and SRS (1993). 

b. Maximum annual missed dose calculated from OCAS-IG-001 (NIOSH 2000). 

E.4.1.5.3 Dosimeter Type

Table E-07 presents the maximum potential missed dose according to SRS facility, time 

period, exchange frequency and MDL.

Table E-07 Missed Photon Dose According to Dosimeter Type

Dosimeter

Type

Period of Use MDL

(mSv)

Exchange 

Frequency

Max. Annual 

Missed Dose

(mSv)

Two 

Element 

Film

Prior to 3/52 through 9/30/1957 0.4 Weekly (n = 52) 20.8

10/1/1957 through 11/8/1959 0.4 Biweekly (n=26) 10.4

MultiElement 

Film

11/9/1959 through 12/31/1964 0.4 Biweekly (n=26) 10.4

1/1/1965 through 12/31/1965 0.4 4-week (n=13) 5.2

1/1/1966 through 3/31/1970 0.4 Monthly (n=12) 4.8

TLD

4/1/1970 through 6/30/1983 0.15 Monthly (n=12) 1.8

7/1/1983 to 2003 (ongoing) 0.05 Monthly (n=12) 0.6

0.05 Quarterly (n = 4) 0.2

E.4.1.5.3 Year

Table E-08 presents the potential maximum missed photon dose according to year 

(actually by period according to dosimeter type and exchange). 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 185 of 188

Table E-08 Missed Photon Dose According to Year

Period of Use(a) MDL 

(mSv)

Exchange Frequency

Max. Annual 

Missed Dose

(mSv)

Prior to 1958 0.4 Weekly (n = 52) 20.8

1959 0.4 Biweekly (n=26) 10.4

1960 through 1964 0.4 Biweekly (n=26) 10.4

1965 0.4 4-week (n=13) 5.2

1966 through 1969 0.4 Monthly (n=12) 4.8

1970 through 1983 0.15 Monthly (n=12) 1.8

1984 to 2003 (ongoing) 0.05 Monthly (n=12) 0.6

(a) Actual periods by mm/dd/yyyy categorized according to year.

E.4.1.5.4 Energy Range

An estimate of the missed photon dose by energy range is possible based on the type of 

facility and predominant radionuclides such as intermediate (>100 keV) energies for all 

facilities handling activation and fission product nuclides, primarily lower (<100 keV) 

energy photons for plutonium facilities and lower energy photons for uranium fuel 

fabrication facilities. The recorded dose from the dosimeter response does not typically 

provide sufficient information to estimate discrete energy ranges. It is possible to 

examine the energy response characteristics of the respective multi-element dosimeters 

but this analysis does not recognize the substantial uncertainties present in the 

workplace associated with shielding, radiation scattering and mixed radiation fields.

E.4.1.6 Missed Neutron Dose Prior to 1971

Prior to 1971, the NTA film used at SRS did not provide a claimant-favorable estimate of 

the neutron dose. Essentially all SRS workers who may have received significant 

neutron dose did receive a measured photon dose. As such, using a ratio of the 

potential neutron dose to the measured photon dose is done as a claimant-favorable 

option to reconstruct an individual worker neutron dose. Table E-08 provides a 

summary of the recommended neutron to photon ratios for the different process areas. 

As can be determined from E-09, the recommended method to apply the ratio is as a log 

normal distribution using the geometric mean and geometric standard deviation. 

Caution, however, should be employed when using Table E-09 and consideration should 

be given to an individual energy employee’s neutron to photon ratio if there is sufficient 

individual data. For example if an individual has recorded neutron dose either prior to or 

after the estimation period that has a higher neutron to photon ratio than what is 

provided in the table, then the higher neutron to photon ratio should be used provided 

that the work assignments were similar for the two time periods. The values in Table E09 should be used when there is known technical limitations in the neutron dosimetry 

technology or no neutron monitoring data. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 186 of 188

Table E-09 Neutron to Photon Ratios by SRS Process

Process Description / Buildings

Operations Neutron to Photon Ratio

Begin End

Geometric 

Mean

(GM)

Geometric 

Standard 

Deviation 

(GSD)

Upper 95th% 

Fuel 

Fabrication

Produced reactor fuel and target assemblies. Intermittent production mixed oxide fuel and plutonium – aluminum 

alloy targets.

(321-M) (a) (a) (a) 

Reactors During Reactor Operation: Low-level neutron exposure through shielding. In addition, some intermittent neutron 

exposure from mixed oxide fuel and targets. 

(105C) 3/1955 1/1987

0.18 2.52 0.82

(105K)

10/1954 1/1988

1992 7/1992

(105L)

7/1954 2/1968

1985 6/1988

(105P) 2/1954 8/1988

(105R) 12/1953 6/1964

Plutonium 

Production

Plutonium Finishing Process: Plutonium enters the process as PuF4 and is then fired into production pucks. Work 

is primarily conducted in glove boxes with predominant close anterior exposure to workers. Radiation levels at the 

beginning of the process are fairly constant while levels at end of process are closely related to production levels. 

Plutonium Finishing Process

(221-H, B-Line)

11/1954 1988 0.91 2.84 5.05

Plutonium Finishing Process

(221-F, B-Line)

10/1955 1988 0.36 2.52 1.65

Plutonium Storage

Production Control Laboratory

(772-F)

1955

Plutonium Fuel Facility

(235-F)

1955

Radionuclide

Production 

and 

Calibration

Radiation characteristics are highly dependent upon the radionuclide being produced. SRS has produced several 

radionuclides including 238Pu, 239Pu, americium, curium, californium, etc. At the calibration facility radiological 

instruments are calibrated using a variety of sources with different neutron energy spectra. 

 

Californium Production Facility (773-

A)

0.62 2.29 2.41

Calibration Facility (736-A) (b) (b) (b)

(a) During Pu-Al target production, workers with potential for neutron exposure wore NTA film, thus LOD values listed in 

table 5.5.2-1 should be used. 

(b) Neutron to photon ratio for the calibration facility is not expected to be greater than the ratio observed in the 

californium production facility. As a claimant favorable approximation the Californium Production facility ratio can be 

used. 

Due to potential for photon missed dose effecting the neutron to photon ratio, when 

applying the ratio in Table E-09, the photon dose should be the combination of both the 

measured photon dose (Dosimeter Dose) and the photon missed dose. When 

overestimating dose reconstructions, the upper 95% ratio can be used as a claimantfavorable worst case ratio and the associated neutron uncertainty can be omitted. In 

underestimating dose reconstructions, the geometric mean can be used as a best case 

ratio. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 187 of 188

E.4.1.7 Missed Neutron Dose After 1970 

Following SRS implementation of the TLND on January 1, 1971 recorded neutron dose 

has been reasonably accurate. As such, OCA-IG-001 guidance on missed dose should 

be followed in accordance with the LOD values presented in Table E-10. 

Table E-10 SRS TLND Limits of Detection 

Dosimeter Type

Dates of Use Exchange 

Frequency

Reported 

LOD

(mrem)

Estimated 

LOD

(mrem)

Max. Annual 

Missed Dose

(mrem) Begin End

SRS Hoy TLND 1/1/1971 12/31/1994 Monthly 

(n=12) 10(c) 20 240

Panasonic TLND 1/1/1995 Present Monthly 

(n=12) 15 15 180

(a) Based on Wilson et al (1990)

(b) Correction for estimated 25% under response of NTA film for some facilities

(c) Minimum Recorded Dose not necessarily LOD

E.4.1.8 Organ Dose Equivalent

Once the adjusted photon and neutron doses have been calculated for each year, the 

values are used to calculate the organ dose distribution for the primary organ of interest 

identified in the claim. Table E-05 summarizes default workplace geometries. These can 

be used in case more applicable values (NIOSH 2002) cannot be determined. A range 

of reasonable estimates may be evaluated to arrive at a claimant-favorable selection.

Table E-11 Default Exposure Geometries to Calculate Organ Dose

Claim Status(a) Job Category (b) Exposure 

Geometry

Percentage (c)

Non-compensable All AP 100%

Compensable - Workers All AP 50%

ROT 50%

Compensable - Supervisors All AP 50%

ISO 50%

(a) Specific time spans for the various SRS facility operation can be found in the section 5.

(b) More than one job category may be needed for longer-term employed workers. 

(c) Apply this percentage to the dose conversion factor (NIOSH 2002a, Appendix B) to arrive at 

the total organ dose equivalent from the adjusted recorded dose. 

E.4.2 Parameter #2 

Parameter #2 is the standard deviation of the normal distribution for the organ dose. 

The individual dose result for each dosimeter exchange period will be available to 

calculate the mean and standard deviation for each year. If not available, the adjusted 

organ dose can be used for each year and a default standard deviation value used for 

parameter #2. 
Effective Date: 07/15/2003 Revision No. 00 Procedure No. 

ORAUT-TKBS-0003

Page 188 of 188

E.4.3 Organ Dose Conversion Factors 

OCAS-IG-001 Appendix A (NIOSH 2002) contains a detailed discussion of the 

conversion of measured doses to organ dose equivalent, and Appendix B contains the 

appropriate dose conversion factors (DCFs) for each organ, radiation type, and energy 

range based on the type of monitoring performed. In some cases, simplifying 

assumptions are appropriate.